Textbook of Radiological Safety.pdf

TEXTBOOK OF

RADIOLOGICAL SAFETY

TEXTBOOK OF

RADIOLOGICAL SAFETY

K Thayalan PhD Professor and Head Radiological Physics Department Barnard Institute of Radiology and Oncology Government General Hospital and Madras Medical College Chennai, India

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Textbook of Radiological Safety © 2010, Jaypee Brothers Medical Publishers (P) Ltd. All rights reserved. No part of this publication should be reproduced, stored in a retrieval system, or transmitted in any form or by any means: electronic, mechanical, photocopying, recording, or otherwise, without the prior written permission of the author and the publisher. This book has been published in good faith that the material provided by author is original. Every effort is made to ensure accuracy of material, but the publisher, printer and author will not be held responsible for any inadvertent error (s). In case of any dispute, all legal matters are to be settled under Delhi jurisdiction only. First Edition: 2010 ISBN 978-81-8448-886-9 Typeset at JPBMP typesetting unit Printed at Ajanta Offset

Dedicated to My Parents (Late) Thiru K Kuppusamy Jayamkondar Thirumathi K Arukkani

Dr B R Ambedkar Institute Rotary Cancer Hospital All India Institute of Medical Sciences Ansari Nagar, New Delhi - 110 029 Tel (Off): 91-11-26594864, 26594798 Fax: 91-11-26589821 E-mail: [email protected] Dr GK Rath MD Professor and Head, Department of Radiation Oncology Chief, DRBRAIRCH, AIIMS Ex President, AROI

Foreword It brings me immense pleasure to write the foreword for this book which is focusing on radiation protection. There was a long-felt necessity for such a textbook. Dr Thalayan has an extensive experience in the field of Medical Physics and this book sums up his vast experience for the benefit of the readers. The author must be complemented for the lucid style of writing. It contains all the essential aspects of radiological safety. The chapter on "Regulations and Dose Limits" is of particular relevance as it contains details of regulatory aspects. The book will go a long way in helping the Radiation Oncology, Nuclear Medicine, Radiology and Medical Physics Community and will be very useful for the health care providers at all levels in these specialties. The chapters are concise and complete in all aspects. Large numbers of illustrations have been included to explain the subject matter. Bibliographies at the end of each chapter have been included to serve as additional reading material on the subject. I wish Dr Thalayan all success in his maiden venture. Prof GK Rath MD Professor and Head, Department of Radiation Oncology Chief, DRBRAIRCH, AIIMS Ex President, AROI

Preface It gives me immense pleasure to come out with a textbook on radiological safety, a unique textbook. It is my long-felt dream to have a complete book, on radiological safety, covering the entire fields of radiology i.e., diagnostic radiology, nuclear medicine and radiotherapy. Radiation is analogous to fire which has both beneficial and harmful effects. The inherent philosophy is to minimize the hazards and maximize the benefits in order to bring down radiation doses within the regulatory control limits by which we can ensure the safety of the occupational workers as well as the patient and public. Hence, it is important that every one should be aware of the safety concepts, dose limits, regulation, waste disposal, etc., to establish a safe work culture while handling radiation sources in the hospital. As such, no single document is available for the above purposes, and the information is collected from safety codes and guides of international and national agencies like IAEA, NCRP and AERB. An attempt has been made to bring all the relevant information including safety terminology, biological effects, exposure control, monitoring, planning of the installation, quality assurance, regulation, personnel safety, transport, waste disposal and radiation emergency, etc. in the form of a book. The whole objective is to remove misconception about radiation and prepare the minds of younger generation to face the future challenge confidently. This book is intended for postgraduates of medical physics, diagnostic radiology, nuclear medicine and radiotherapy. This book may also find a place for the preparation of RSO examinations for medical physicists. Moreover, an attempt has been made to bring quantitative data from international reports and recommendations, wherever necessary, with practical examples and illustrations. This may enable the new entrants to plan a radiation facility, carry out quality assurance and radiation survey without much cumbersome and perform the day-to-day medical physicist’s job with ease and involvement. Large numbers of tables and figures are incorporated wherever necessary for better understanding of the reader. I am very much thankful to my family members for their support and cooperation. I also acknowledge the assistance offered by the Dr Kamakshi Memorial Hospital staff, especially the medical physics colleagues, in the preparation of the text. I also thank Mrs G Shakunthala for neatly typing the manuscript. I also acknowledge my teachers, by whom I got inspiration and passion towards teaching. I invite the readers to offer constructive comments for the future improvement of the book. K Thayalan

Contents 1.

Safety Concepts ........................................................................................ 1 Introduction 1 Radiation units 2 Equivalent dose 4 Effective dose or effective dose equivalent 4 Committed dose 6 Collective dose 6 Genetically significant dose 7 Detriment 8 Annual limit on intake 8 ALARA 8 Sources of radiation 9

2.

Biological Effects of Radiation ............................................................. 14 Cell 14 Interaction of radiation with tissue 14 Linear energy transfer 16 Biologic effects 17 Radiation effects on DNA 22 Radiation effects in utero 23 Radiation risk 24 Ten day rule and its present status 29

3.

Radiation Exposure Control .................................................................. 31 Time 31 Distance 32 Shielding 34 Half value layer 35 Sources of exposure 37 Leakage limits 39 Protective barrier design 40 Facility design for diagnostic X-rays 42 Facility design for nuclear medicine 47 Facility design for radiotherapy 48

4.

Planning of Radiological Facility ......................................................... 64 General guidelines 64 Establishing a diagnostic X-ray facility 65 General radiography installation 68 Fluoroscopy installation 68 Mammography installation 69 Computed tomography installation 70 Establishing a nuclear medicine facility 71 In-vivo diagnostic facility 74 In-vitro and radioimmunoassay (RIA) 75

Textbook of Radiological Safety Radionuclide therapy 78 Establishing a radiotherapy facility 79 Brachytherapy facility design 91

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5.

Radiation Monitoring ............................................................................. 95 Personnel monitoring 95 Film badge 96 Thermoluminescent dosimeter 97 Pocket dosimeter 100 Personnel monitoring systems and features 102 Area monitoring 103 Radiation survey in diagnostic radiology 107 Radiation survey in nuclear medicine 111 Radiation survey in radiotherapy 112 Calibration and maintenance of radiation monitoring instruments 117

6.

Quality Assurance ................................................................................. 119 Introduction 119 Quality assurance for diagnostic radiology 119 Quality assurance for radiography unit 120 QA for mammography X-ray unit 133 QA for fluoroscopy X-ray unit 134 Quality assurance for computed tomography 134 Quality assurance for nuclear medicine 137 QA for gamma camera 138 QA for single photon emission computed tomography (SPECT) 140 Quality assurance for PET-CT 141 Image quality tests 146 QA for radiopharmaceuticals 147 Quality assurance for radiotherapy 154 QA for linear accelerator 154 QA for HDR brachytherapy 159

7.

Regulations and Dose Limits .............................................................. 167 Atomic energy act-1962 167 Atomic energy regulatory board 167 Radiation protection rules-2004 168 Regulatory controls for diagnostic X-ray equipment and installations 181 Regulatory controls for nuclear medicine facilities 184 Regulatory control for radiotherapy equipment and installations 190

8.

Personnel Protection ............................................................................. 204 Radiography 204 Protection in fluoroscopy 212 Protection in computed tomography 214 Protection in pediatric imaging 215 Pregnancy and radiation 221 Protection in nuclear imaging 227 Protection in radionuclide therapy 232 Pregnancy and radiation protection in nuclear medicine 233

Contents Radioiodine therapy and pregnancy 235 Staff protection 237 Personnel safety during source transfer operations of teletherapy and HDR brachytherapy units 238 Pregnancy and radiation protection in radiotherapy 240 Records 243 9.

Transport of Radioactive Materials ................................................... 245 Introduction 245 Types of packages 246 Transport index 248 Packaging and package requirements 249 Preparation of the package for transport 253 Marking of the package 253 Labeling of the package 254 Placards 256 Booking, storage, transport and delivery of package 257 Consignor’s declaration 258 Tremcard 259 Information to carriers 260

10.

Radioactive Waste Disposal ................................................................ 267 Introduction 267 Waste management 267 Sources and nature of waste 268 Classification of waste 269 Types of radioactive waste 270 Disposal of low activity wastes into the environment 276 Disposal of radioactive effluent into the ground 276 Disposal of P-32 and I-131 into municipal sewers by medical users 277 Disposal of radioactive waste from nuclear medicine procedures 278 Routine protective clothing 280 Decontamination procedures 281

11.

Radiation Emergencies ......................................................................... 289 Type of radiation accidents 289 Diagnostic radiology-skin injuries 294 Nuclear medicine: Radiation accidents 297 Radiotherapy: Radiation emergencies 300 Brachytherapy: Radiation accidents 308 Emergency preparedness: Actions 309 Medical management of personnel exposed to radiation 313 Index .......................................................................................................... 319

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Chapter

1

Safety Concepts

INTRODUCTION Radiation is small pockets of energy, which travels as waves and transfer energy from one point to another point. There are two types of radiation namely: (i) photons, e.g. X,γ, and (ii) particles, e.g. e, p, n and α. Radiation is a double edged weapon, analogous to fire, which possess both benefits and hazards. Radiation hazards were witnessed by the following events in early days: 1. Uranium mine workers 2. Watch dial painters-switzerland 3. Atomic bomb explosion-Hiroshima and Nagasaki 4. Radiobiology experiments. Radiation was used in medicine immediately after the discovery of X-rays by W.C. Roentgen on November 15, 1895. It was used in India in 1898 within 3 years of its discovery. The Indian army employed hefty porters to carry the cargo,(100 bounds) on a pole for 200 miles in the Hyber pass region (Pakistan). That cargo had the first X-ray tube used in India. Major Bewoor used it effectively in the North-West frontiers. The civilian use of X-rays in India began in 1900 at the Government General Hospital, presently the Barnard institute of Radiology and Oncology, Chennai. Radiation hazards were realized in the beginning of 20th century. The X-rays were used indiscriminately in the early years and have caused visible damage to several physicians and X-ray enthusiasts. Within 6 months of their use, several cases of erythema, dermatitis and alopecia were reported among X-ray operators and their patients. In 1902 the first X-ray induced skin cancer was reported. In 1921 Ironside Bruce, a pioneering radiologist in a London Hospital died of cancer at the age of 38. Similarly several lives were lost due to excessive X-ray exposures. In 1915, the British Roentgen Society made the first radiation protection recommendations. To regulate the safe use of radiation the “British X-ray and Radium protection committee” was formed (1921). It was made as an International Committee in 1928 and later (1950) transformed as “International Commission on Radiological Protection”(ICRP). The ICRP is the first standard setting body formed, for the purpose of radiological safety. The similar organization at the USA is the National council on radiation protection and measurements (NCRP), which was formed in 1946.

Textbook of Radiological Safety These bodies issue periodical reports on radiation safety aspects of various application of ionizing radiation. These concepts are explained in the following paragraphs and chapters. RADIATION UNITS Exposure – Roentgen The term exposure (X) refers the radiation quantity measured in terms of ionization in air, in a small volume around a point. Exposure is a source related term. Exposure from an X-ray source obeys inverse square law. The unit of exposure is roentgen (R). One roentgen shall be the quantity of x or gamma radiation such that the associated corpuscular emission per 0.001293 grams of air (1cc of dry air at NTP), produces in air, ions carrying 1 e.s.u. of quantity of electricity of either sign. The unit may also be defined in terms of SI unit as 1R = 2.58 × 10-4 C / kg of air There are some difficulties in the unit of roentgen. It is not a unit of dose, which is a measure of absorbed energy. It can be used only up to a photon energy of 3 MeV. It is defined only for x and gamma radiation in air. Kerma Kerma stands for kinetic energy released in the medium, which describes the initial interaction of the photon with an atom, that take place in the medium. When radiation interacts with matter, the uncharged particles (photons and neutrons) transfer kinetic energy to the charged particles (e and P). Kerma (K) is the measure of kinetic energy transferred to the charged particles. It is defined as the sum of the initial kinetic energy of all the charged ionizing particles, liberated by photons in a material of unit mass. The unit for kerma is Joul per kilogram (J/ kg). The SI unit is Gray and the special unit is rad. Absorbed Dose—Rad / Gray The term absorbed dose (D) refers the amount of energy absorbed per unit mass of the substance. The unit of absorbed dose is rad (r), which means radiation absorbed dose. 1 rad =100 ergs/gram. This unit is independent of type of radiation and the medium. The SI unit of absorbed dose is Gray (Gy). 1Gy = 1 J / kg The unit rad is related to gray as 1Gy = 100 rads.

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Safety Concepts Exposure Rate Constant It is defined as the exposure per hour from 1 mCi point source at a distance of 1cm and it is expressed in R-cm2/mCi-h. For example, the exposure rate constant of cobalt, cesium and iridium radioisotopes are 13.07, 3.26 and 4.69 respectively. RHM and RMM If the exposure rate constant is defined for 1Ci source at 1m, then it is called RHM (Roentgen Hour Meter). It is defined as the exposure per hour from 1Ci point source at a distance of 1m and it is expressed in R-m2/Ci-h. In practice RHM of a given radioisotope can be obtained by dividing the exposure rate constant by a factor 10. Hence the RHM of cobalt, cesium and iridium radioisotopes are 1.307, 0.326, and 0.469 respectively. Mostly RHM is employed for calibration purposes in Brachytherapy and industrial radiography. Instead of hour one can also express it for one minute, then it is called RMM (Roentgen Minute Meter). It is defined as the exposure per minute from 1 Ci point source at a distance of 1m and it is expressed in R-m2/Cimin. For example, if the RHM of the cobalt radioisotope is 1.307, the corresponding RMM value is 1.307/60 = 0.0217. RMM is the most preferred and useful terminology in teletherapy source calibration. RMM and CURIE The unit of activity is curie and it is defined as the number of disintegration per second from 1 gram of radium and it is found to be 3.7 × 1010. In the case of cobalt, 0.0217 RMM corresponds to 1Ci, hence 1RMM = (1/0.0217)= 46.08 Ci. Similarly, it can be applied to different radioisotopes (Table 1.1). Table 1.1: RHM and RMM of different radioisotopes Radioisotope

RHM

RMM

Activity (Ci) equivalent to 1 RMM

Radium-226 Cobalt-60 Cesium-137 Iridium-192

0.825 1.307 0.326 0.469

0.0137 0.0217 0.0054 0.0078

72.99 46.08 185.18 128.20

Example 1: If a Cobalt teletherapy source is purchased with 200 RMM capacity, what is the corresponding activity of the source in Ci? 1 RMM = 46.08 Ci 200 RMM = 46.08 × 200 = 9216.58 Ci

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Textbook of Radiological Safety Example 2: If a Cobalt teletherapy source is purchased with 10,000 Ci capacity, what is the corresponding RMM value? 46.08 Ci = 1RMM 10,000 Ci = (1/46.08) × 10,000 = 217 RMM. EQUIVALENT DOSE The biological effects of radiation depend not only on absorbed dose (D) but also on the type of radiation. Hence, the ICRP report 26 (1977) introduced the dosimetric quantity Equivalent dose (HT). It is the absorbed dose averaged over a tissue or organ and weighted for the radiation quality that is of interest, and is given as HT=D × WR Where WR is the weighting factor for the radiation type and it is analogous to RBE in radiobiology. Earlier the term quality factor (Q) was used to evolve dose equivalent (absorbed dose × quality factor). This is discontinued now and replaced with equivalent dose, which is an average dose and not a point dose. Table 1.2 gives the suitable weighting factors for various type of radiations. Table: 1.2: Radiation weighting factors (WR) Radiation type

WR

Radiation type

WR

Photons (all energies) Electrons Neutrons,<10keV Neutrons,10keV-100keV Neutrons,100keV-2MeV

1 1 5 10 20

Neutrons,2MeV-20MeV Neutrons >20MeV Protons Alpha particles

10 5 5 20

Sievert (Rolf Sievert, Swedish Radiologist) is the SI unit of equivalent dose and one Sievert (Sv) =1 Joule/kilogram. Also 1 Sv=100 Rem (Radiation equivalent men), where Rem is the special unit of equivalent dose. In practice, milli sievert (mSv) unit is used. 1 Sv = 1000 mSv 1 mSv = 100 mRem. EFFECTIVE DOSE OR EFFECTIVE DOSE EQUIVALENT The whole body exposures are not uniform and dose equivalents for various tissues may differ markedly. Hence, the radiation induced effects vary with the sensitivity of the organ. To account these non uniform irradiation and organ sensitivity variation, the ICRP-26 introduced the term effective dose (E), which describes the dose to the whole body and is derived from 4 equivalent dose. It is defined as

Safety Concepts E = ΣWT × HT where WT is the weighting factor for the tissue T, HT is the mean equivalent dose received by the tissue and E is the summed organ or tissue doses as an overall whole body dose. This quantity expresses the overall measure of health detriment associated with each irradiated tissue or organ as a whole body dose and considers the radiosensitivity of each irradiated organ or tissue. It is used to evaluate the probability of stochastic effects at low doses. The weighting factor of a particular tissue or organ is the risk of stochastic effects being induced in the organ when singly irradiated, compared to the total risk of inducing stochastic effects if the same radiation dose is received by the whole body. The Table 1.3 gives the tissue weighting factors for various tissues. It is seen that testes and ovaries are the most radiosensitive tissues as they have the highest value of weighting factor as per ICRP 60. However, the radiosensitivity of the breast tissue and gonads are reassessed by the ICRP (2005) and the revised weighting factors are given the table. Organ of higher sensitivity carries a higher risk for a given dose. The sum of the weighting factors is unity. The unit of effective dose is Sievert (Sv). Table: 1.3: Tissue weighting factors (WT) Tissue

WT Tissue (ICRP 60)

Testes and ovaries(Gonads) Red bone marrow Colon, lung and stomach Breast, urinary bladder Thyroid, liver, esophagus Bone surfaces Remainder

0.20 0.12 0.12 0.05 0.05 0.02 0.05

Bone marrow, breast Colon, lung, stomach Bladder, esophagus Gonads, liver, thyroid Bone surface, brain, kidneys Salivary glands, skin Remainder

WT (ICRP 2005) 0.12 0.12 0.05 0.05 0.01 0.01 0.10

Example 3: In a CT scan study the tissues breast, lung, bone marrow and thyroid receive dose of 21, 23.5, 5.17 and 2.30 mSv respectively. Calculate the effective dose with both old and revised tissue weighting factors. The effective dose: E = ΣWT × HT = (21 × 0.05+ 23.5 × 0.12+5.17 × 0.12+2.3 × 0.05) = 4.60 mSv(ICRP 60) = (21 × 0.12+ 23.5 × 0.12+5.17 × 0.12+2.3 × 0.05) = 6.07 mSv (ICRP2005) The revised tissue weighting factors predicts 32 % higher risk for a given radiation dose.

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Textbook of Radiological Safety COMMITTED DOSE If an individual is subjected to a radiation burden over a period of time, then committed dose is the term to be used. It is the absorbed dose the individual receives as a result of the intake of radioactive material. The individual will continue to receive a dose of radiation as long as the traces of radioactivity remain with in the body. The factor which determines the remaining activity in the body is the effective half life (T1/2 eff ). It is related to the physical half life (T1/2 phys ) and biological half life (T1/2 biol) as follows: 1 1 1 = + T1/2 eff T1/2 phys T1/2 biol

One can not alter the physical half life, which is a character of a given radionuclide. Whereas the biological half life can be reduced by increasing the rate of excretion of the radionuclide from the body. For example, a radio nuclide has a physical half-life of 6 hours and a biological half-life of 3 hours, then 1/ T1/2 eff = (1/6) + (1/ 3), and T1/2 eff = 2 hours. The effective half-life is always less than either the physical or biological half-life The committed dose equivalent is the quantitative assessment of the effect of a particular intake of radioactivity over the whole of a individual’s working life. It is defined as the dose equivalent accumulated over a period of 50 years following the intake of radioactive material. In the case of children the period is taken as 70 years. It is defined HT(t) = HT × t where t is the period of time in years. If the committed organ or tissue equivalent dose is multiplied by the suitable tissue weighting factors then the sum of the products is called committed effective dose (E(t)). E(t) = ΣHT(t) × WT The other factors which influences the dose equivalent are; (i) the concentration of the activity in the organ,(ii) whether the concentration is uniform or localized, (iii) decay system, (iv) radiation weighting factor, (v) size and shape of the organ,(vi) proximity of other organs and (vii) weighting factor of the organ. COLLECTIVE DOSE

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To assess the overall effect of radiation dose on a large group of people, the individual dose may be multiplied by the population number exposed and it is called the collective dose. If N is the number of population receiving a mean organ equivalent dose HT , over a period of time t, then the collective equivalent dose (ST) is given by

Safety Concepts ST = Σ HT(t) × N Collective Effective Dose Equivalent In a similar way the collective effective dose (S) can be defined. It is the whole body exposure to a population group exposed to radioactive materials in the environment and can cover successive generations of the populations being studied. S = E(t) × N In a country, if a population of 10 million people are exposed to a background dose of 3 mSv, then the collective effective dose is 30,000 manSv. The collective effective dose equivalent (CEDE), can be used as a method to assess the impact of human health from population radiation exposures and it is expressed in person-Sv. The CEDE values for different occupations are given in Table 1.4 Table 1.4: Annual collective effective dose equivalents for workers in various occupations in USA Occupation

Annual collective effective dose equivalents (person-Sv)

Nuclear power operation Medicine Industry Airline crews and attendants Uranium miners

550 420 390 170 110

The uranium miners CEDE is small due to relatively small size of the workforce involved in the occupation. The total annual CEDE for all occupationally exposed workers is about 2000 Person–Sv, for all occupations. Whereas the annual CEDE attributable to natural background radiation for the same population is about 3200 person-Sv. It means that the occupational workers getting an additional 63% (2000/3200) over the natural background. In general the CEDE is decreasing in medicine due to small size of the exposed population and improved health physics practices. GENETICALLY SIGNIFICANT DOSE The genetically significant dose (GSD) is defined as that equivalent dose that, if received by every member of the population, would be expected to produce the same genetic injury to the population as do the actual doses received by the irradiated individuals. It is expressed in Sievert (Sv). It is 7

Textbook of Radiological Safety used to assess the genetic risk or detriment to the whole population from radiation exposure, especially in medicine. It assumes a linear dose effect relationship. For example, the patients undergo X-ray examinations may receive a dose of about 10.0 mGy. If the same dose was received by every member of the population, it would be expected to produce the same total genetic effect on the population. The GSD accounts the child bearing potential of the patient population. The genetically significant dose (GSD) is used to assess the genetic risk or detriment to the whole population from radiation exposure. In this the equivalent dose to the gonads of each exposed individual is weighted for the number of progeny expected for a person of that sex and age. Annual GSD from all radiation sources is about 1.3 mSv that includes 1.02 mSv (78%) from natural background and 0.28 mSv (22%) from technological sources (NCRP-report 93,1987). Among the technological sources, the major contributor is diagnostic X-rays, 0.20 mSv (15%). Among the 15 %, onethird is attributable to male and two-third to females. The higher proportion of female component is due to the location of the ovaries within the pelvis, which places them in the primary beam during most abdomino pelvic examinations. DETRIMENT Detriment is a measure of harm caused by exposure to radiation. It is the expectation of harm incurred from an exposure to radiation, taking into account not only the probability of each type of deleterious effect, but also the severity of the effect. Usually several parameters (e.g, probability of death and reduction of life expectancy) are considered to arrive the mean health detriment. Health detriment is an estimate of the risk of reducing in length and quality of life occurring in a population following exposure to ionizing radiations. ANNUAL LIMIT ON INTAKE The Annual limit on intake (ALI) is the that quantity of radionuclide which, taken into the body during 1 year, would lead to a committed effective dose equal to the occupational annual limit on effective dose. ALARA As Low As Reasonably Achievable (ALARA) term was introduced by ICRP26. It states that doses to patients and staff should be kept as low as reasonably achievable. Every reasonable effort must be made to reduce radiation levels below the stated dose limits within economic and social limits.

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Safety Concepts SOURCES OF RADIATION The sources of radiation are classified into (i) Natural radiation sources, (ii) Enhanced natural sources, (iii) Artificial radiation sources (man made) and (iv) Occupational exposures. The annual average per capita total effective dose equivalent is 3.6 mSv (NCRP-93,US data). About 82% of the above exposure (3 mSv) arise from naturally occurring sources, 18% (0.6 mSv) arise from technologic enhancements of naturally occurring sources and artificial radiation sources (diagnostic X-ray is the major contributor). Background radiation involves both natural and man made low level radiation exposure to all members of the public. This will vary with region and Kerala and Brazil have high background levels of radiation (100 mSv/year). Natural Radiation Source The natural radiation sources includes (i) Cosmic rays,(ii) Terrestrial (Primordial) radionuclides, and (iii) Internal Radioisotopes. Cosmic Rays Cosmic rays are extraterrestrial radiation that strikes the earth’s atmosphere, that includes primary and secondary. Primary cosmic rays, in which protons accounts for 80%. The primary cosmic rays collide with atmosphere, producing showers of secondary particles (electrons, muons) and electromagnetic radiations. The average per capita equivalent dose is 270 μSv per year, which makes 8% of the natural background. Cosmic exposures increase with altitudes. It is estimated that at 30,000 ft altitude the equivalent dose is about 5 mSv per hour and it is doubling in every 1500 feet. It is greater at the earth poles than the equator. Structures provide some protection against cosmic rays, and hence the indoor effective dose is 20% lesser than outdoor. Air travel increases individual’s cosmic ray exposures. Air crews and frequent fliers receive an additional annual equivalent dose of 1mSv. A 5 hr transcontinental jet aircraft travel result in 25 μSv equivalent dose. Apollo astronauts received an average equivalent dose of 2.75 mSv during the lunar mission. A part of secondary cosmic ray particles collide with stable atmospheric nuclei and produces cosmogenic radionuclides; e.g.147 N (n,p) 14 C, but their contribution to natural background is very little. 6 Terrestrial Radiations Terrestrial radionuclides that have been present on earth since its formation are called primordial radionuclides. Their physical half lives are comparable to the age of the earth (4.5 billion yeras). Their decay products are the major contributors of terrestrial radiations. They mainly contribute in the form of 9 external exposure, inhalation, and ingestion.

Textbook of Radiological Safety External exposure: K-40,U-238, and Th-232 are mainly responsible for external exposure and they account an equivalent dose of 280 μSv per year. This may vary depending upon the local concentration of terrestrial radionuclides. Inhalation: Rn-222 (U-238) is a noble gas, decays to polonium-218 by alpha emission with half life of 3.8 days. Its decay products are the most significant source of inhalation exposure. It is deposited in the tracheobronchial region of the lung. Radon concentration vary widely both seasonal and diurnal. It emanates from sail and is restricted by structures. Weatherproofing of homes, energy conservation techniques, decreased ventilation are resulting in higher indoor radon concentration. Radon inhalation accounts an equivalent dose of 2 mSv/year to the bronchial epithelium. It accounts for about 55% of natural background, which can be easily measured and reduced. Ingestion: Ingestion of food and water is the second largest source of natural background in which K-40 is the most significant. It is a naturally occurring isotope of potassium having higher concentration at the skeletal muscle. It accounts an average equivalent dose rate of 400 μSv/year. Internal Radionuclides Internal radionuclides includes K-40 and C-14, which are present the in the human body. The main contributor is K-40,which emits β and γ rays and decays with a half life of 1.3 × 109 years. Enhanced Natural Sources Enhanced natural sources mainly consists of consumer products. The largest contributor is tobacco products, which burdens the bronchial epithelium. It produce an effective dose equivalent of 2.8μSv/year. Radon gas dissolved in domestic water supply can contribute 10-60 μSv/ year. Building materials consists of uranium, thorium and potassium and these are present in brick, concrete, granite which may contribute an annual effective dose of 30 μSv /year. Mining and agricultural activity contribute to a lesser level by fertilizers (uranium, thorium decay products and K-40). Cumbustible fuels including coal, natural gas and consumer products includes smoke alarms (americium-241), gas lantern mantles (thorium), dental prostheses, certain ceramics, optical lenses (uranium) contribute <1% annual effective dose Artificial Sources The artificial sources of radiation includes medical exposure, radioactive

10 fallout, nuclear power and occupational exposure.

Safety Concepts Medical Exposure The majority of the exposure is from medical X-rays (Fluoroscopy & Computed tomography) which contribute to 58% of the artificial radiation exposure. Next contributor is the nuclear medicine which is 21%. Both produces an annual average effective dose equivalent of 540 μSv per year. It accounts for about 69% of artificial radiation. Consumer Products It accounts to 16% of the artificial radiation exposure. Substances in consumer products such as tobacco, the domestic water supply, building materials, and to a lesser extent, smoke detectors, televisions, and computer screens, account for the above exposures. Radioactive Fallout It arises from atmospheric testing of nuclear weapons and consists of Carbon-14 (70%) and other radionuclides including H-3, Mn-54, Cs-136, 137, Ba-140, Ce-144, plutonium and transplutonium elements. It results in an annual effective dose equivalent of <10 μSv. It contributes 2% of the manmade radiation exposures. Nuclear Fuel Cycle The contribution from nuclear power production is very minimal, which is about 1% of artificial radiation (Annual effective dose is <0.5 μSv). It involve all phases of fuel cycle; mining, manufacturing, reactor operations, and waste disposal. The most significant contributor is Carbon-14. Occupational Exposure The occupational exposures associates with uranium mining (12 mSv per year), nuclear power operations, medical diagnosis and therapy, aviation and research, non uranium mining, and application of phosphate fertilizers. It contributes about 2 % of the artificial radiation exposure. Radiologist, X-ray technologist receive an average annual effective dose of 1 μSv. However special procedures involving fluoroscopy and ciniradiology (e.g. cardiac catheterization) may exceed 15 mSv. These are only partial body exposures (head & extremities), if lead apron is used during the procedure. The occupational exposures for various categories are listed in Table 1.5. The UNSCEAR and the NCRP have published the global annual dose contribution from various sources of radiations, which is presented in Table 1.6. The average annual effective dose equivalent to a population from all radiation sources, is obtained by dividing the annual collective effective dose equivalent by the size of the population. The Fig. 1.1 shows the % contribution of various radiation sources to the total average effective dose 11

Textbook of Radiological Safety Table 1.5: Annual occupational effective dose equivalent (US Data) Category

Average annual total effective dose equivalent (mSv)

Uranium miners Nuclear power operations Airline crews X-ray and Nuclear Medicine technologists Radiologists

12.0 6.0 1.7 1.0 0.7

equivalent in the US population, which is estimated as 3.0 mSv/year, excluding smoking. Smoking is the largest contributor (45%) to the average population effective dose equivalent. Hence, lung is the organ that receives highest dose equivalent from both smokers (polonium-210) and non smokers (radon-222). The Fig. 1.2 highlights the % contribution of various types of manmade radiation sources. Table 1.6: Global annual radiation dose distribution from various sources of radiation Sources Radon Gamma rays Cosmic rays Medical Internal Radioactive fall out Occupational discharges Products Total

12

UNSCEAR (1993), Percentage and mSv/year 48.3% 17.1 14.5 11.2 8.6 0.3 <0.1 <0.1 <0.1 100

1.30 0.46 0.39 0.3 0.23 0.007 0.002 0.001 0.0005 2.69 mSv/year

NCRP-93(1987), Percentage and mSv/year 55% 8 8 15 11 <0.3 0.3 0.1 3.0 100

2.0 0.28 0.27 0.53 0.40 0.01

⎫ ⎬ ⎭ 0.07

3.6 mSv/year

Fig. 1.1: The % contribution of various types of man-made sources to the total average effective dose equivalent to the US population

Safety Concepts

Fig.1.2: The % contribution of various components of man-made radiation sources (Source BEIR VII Report)

Medical use of radiation warrant an optimal compromise between clinical utility and radiation dose to patients, staff and the public. The success of radiation protection mainly lies on the education of staff. They should be educated on radiobiology, genetics, risk analysis, methods of reducing radiation dose, decay patterns of radioactivity released in the environment and absorbing power of different materials to different radiation. BIBLIOGRAPHY 1. David JD, Patrick AK, Eugene RJ. The physics of diagnostic imaging (2nd edn.) Hodder Arnold, UK 2006. 2. Donald TG, Paul C, Martin V. Principles of Radiological physics, (5th edn.) Churchill Livingstone 2007. 3. International Commission on Radiological Protection ICRP 26 Annals of ICRP. Pergamon press 1997. 4. Jerrold TB et al. The essential physics of medical imaging, (2nd edn.) Lippincott Williams & Wilkins 2002. 5. Thayalan K. Basic radiological physics. Jaypee brothers Medical publishers (P) LTD, New Delhi 2001.

13

Chapter

2

Biological Effects of Radiation

Radiation biology is a scientific discipline which deals with the study of the action of ionizing radiation on healthy and diseased tissue. It makes us to understand the sequence of events (damage or repair) that occurs after the absorption of radiation energy. There are many variables that determine the biological response to radiation exposure. These include the dose, dose rate, fractionation, nature of exposure (whole body/partial body), cell cycle, presence of radioprotectors /sensitizers and age of the exposed individual etc. CELL Cell is the basic unit of life. Living organisms are made up of either a single cell or many cells. Human being is a multi-cellular organism built up of 10 14 cells. Cell consists of a nucleus, which is surrounded by a viscous liquid known as cytoplasm. Both the cell and cytoplasm are enveloped by a membrane, which is known as cell membrane or plasma membrane. The constituents of cytoplasm control the functions of the cell. The nucleus contains tiny thread like structures known as chromosomes, which are made up of deoxyribonucleic acid (DNA) and protein. The DNA molecules contain all the information required for the cellular function in coded form and thus control the nature and growth of the individual. Sections of chromosomes, which contain information for specific functions, are called genes. Cells of similar nature constitute a tissue and different tissues form an organ. Many organs constitute a system (like respiratory system, digestive system, hemopoietic system and nervous system). Cells are mainly classified into two categories: (i) somatic cells, and (ii) reproductive (or germ) cells. Somatic cells constitute various tissues such as skin, liver, brain etc. Germ cells are those, which participate in the reproductive process. They are sperm in males and ovum in females. All somatic cells in human body contain 46 chromosomes as 22 pairs and two sex determining chromosomes. Reproductive cells contains only 23 chromosomes. INTERACTION OF RADIATION WITH TISSUE Radiation deposits energy in tissues randomly and rapidly (<10-10Sec) via excitation, ionization and thermal heating, in turn produces moving

Biological Effects of Radiation electrons. These electrons interact with atoms and molecules leading to chemical and molecular changes. These changes may appear as biological effects such as chromosome breakage, cell death, oncogenic transformation and acute radiation sickness. These effects appear after a period of time (latent period), which may vary from minutes to years. Major portion of the radiation energy appear as heat, with little biologic significance. Radiation interactions that produce biologic effects are classified as direct and indirect action. In direct interaction, the radiation ionize or excite the molecules such as DNA, RNA and protein directly. It involves rupture of cell membrane and break of chromosome structure, resulting in DNA strands break. The fragments of chromosomes produced in a direct interaction, can join together to form chromosomes with abnormal structures. This is known as chromosomal aberration. The frequency of chromosomal aberrations increases with the radiation dose and hence the magnitude of aberrations is a biological indicator of radiation dose absorbed in human body. Chromosomal aberration analysis (CAA) is useful in determining the radiation dose received by a person who is accidentally exposed to high radiation dose (>100 mGy). In indirect mode, the radiation interacts with the medium and produce radicals which in turn interact with the target molecule. For example, radiation interacts with oxygen and water molecules present in the cell. These interactions produce a large number of free radicals, which are uncharged atoms or molecules with an unpaired electron and hence are highly reactive. Human body tissue is composed of 70-85% water, and the major interaction (three-fourth) is indirect action. The X and γ rays effects in macromolecules of living system are mainly due to indirect interactions. The absorption of radiation by a water molecule (radiolysis) results in ion pairs (H2O+, H2O –),which are unstable and forms free radicals H* (hydrogen) and OH* (hydroxyl) as follows: H2O → H2O+ + H2O – (ion pairs) H2O+ → H+ + OH* H2O– → H*+ OH– OH* + OH* = H2O2 (hydrogen peroxide) H* + O2 = HO2* (Hydroperoxyl radical) Free radicals are extremely reactive chemical species and perform variety of reactions. They act as strong oxidizing or reducing agents by combining with macromolecules. For example, free radicals interfere with cell functions and may inactivate cellular mechanism or break DNA bonds. The damaging effects of free radicals is enhanced with presence of oxygen. In the case of 15 low LET radiation, free radicals are the primary agents that cause the

Textbook of Radiological Safety biologic effects. Approximately two-third of all radiation induced damage is considered to be caused by the hydroxyl free radical. Repair mechanisms exists within cells which repair the cells and return to pre irradiated state. For example DNA single strand break, base damage can be repaired, by specific endonucleases and exonucleases that are present. The Fig. 2.1 presents the physical and biologic responses to ionizing radiation.

Fig. 2.1: Interaction of radiation in cell: Physical and biological response of ionizing radiation

Cells are more radiosensitive during the phase of cell division known as mitosis, when the chromatin material is being distributed to daughter cells. Hence rapidly dividing cells like intestinal epithelium, bone marrow cells, reproductive cells are more radiosensitive. Highly differentiated tissues like muscle and brain are least radiosensitive. Direct and indirect interactions of radiation with cell ultimately result in (i) cell modifications like gene mutation, cell transformation and chromosome aberrations and (ii) cell death. LINEAR ENERGY TRANSFER The linear energy transfer (LET) is a parameter that describes the average energy deposition per unit path length of the incident radiation and it is expressed in keV/μm. All radiations are capable of producing same type of biologic effects, but the magnitude of the effect per unit dose differs. In other words different radiations of equal dose do not produce the same level of biologic response. To evaluate the effectiveness of different radiations the term Relative biological effectiveness (RBE) was introduced and it is defined as follows:

16

RBE =

Dose of 250 kVp X-rays required to produce certain effect Dose of test radiation required to produce the same efffect

Biological Effects of Radiation These effects or endpoint includes chromosomal mutation, cataract formation and acute lethality etc. The relationship between RBE and LET is shown in Fig. 2.2. The RBE is proportional to LET at the beginning, which is relevant for low LET radiations (X, γ, e). Later it increases with LET, suggesting deposition of higher energy in the tissue, which is applicable to high LET radiations(α). Beyond 100 keV/μm, the RBE decreases with increasing LET, due to overkill. This means that radiation deposits excess energy than that necessary to kill the cell. RBE also depends on the total dose and dose rate of the radiation. The LET of the X and γ rays and tissue penetration are listed in Table 2.1 below.

Fig. 2.2: The relation between LET and RBE Table 2.1: The LET of X and γ rays and tissue penetration in HVL X and γ rays Energy (keV)

LET(keV/μm)

HVL(mm) in Tissue

Source

80 120 140 364 511 1000

1.0 1.4 1.5 2.8 3.5 5.2

38 43 46 65 70 100

Diagnostic X-rays Co-57 Tc-99m I-131 Positron emitters Co-60

BIOLOGIC EFFECTS The harmful effects of radiation in human body are classified as (i) somatic effects and (ii) Genetic effects. The radiation effects, arises due to the damage of the somatic cell are called somatic effects. It is produced in an exposed 17

Textbook of Radiological Safety individual during his life time. The magnitude of the somatic effects vary with nature of exposure (whole body or partial exposure). The hereditary effects are due to damage to reproductive cells and manifest in the progeny of the exposed person. Early Somatic Effects (Whole Body Irradiation) Somatic effects may appear immediately after exposure, within a few hours to weeks or much later (after years - decades). The early effects are due to an acute exposure (large doses over a short period of time) and attributed to depletion of cell population due to cell death. Table 2.2 shows the early somatic effects due to the acute whole body exposure to low LET radiation. The amount of radiation damage depends on the rate at which the radiation is delivered. High dose delivered in a short time may result in severe damages to tissues. The same dose delivered over several months allows the repair mechanisms to function fully. Hence, somatic effects listed in the table may not manifest, if the person receives the dose over a prolonged period. Table 2.2: Early somatic effects due to acute whole body exposure to low LET radiation* Dose range

Effect

Less than 0.1 Gy Above 0.1 Gy

No detectable damage Chromosome aberrations detectable(1-2 dicentrics in 500 cells) Above effect + transient reduction in WBC count: emporary sterility in males Above + radiation sickness like loss of appetite, nausea, vomiting , diarrhea (NVD) : complete recovery possible at low doses Severity of above effects increases + damage to blood forming organs (bone marrow, spleen, lymph node): death in 4-8 weeks possible (> 10%) All the above with increased severity + Anemia, infection, high fever (bone marrow syndrome), about 50% of the exposed persons may die within 60 days (LD 50/60). Severity of above effects increases + cells in the gastrointestinal system get severely damaged leading to gastrointestinal syndrome (GIS) like diarrhea, weight loss and fever. Death may occur in 1-2 weeks (100 %). All the above + manifestation of skin damage All the above + severe depression, fatigue, delirium, coma etc. (Central nervous system syndrome[CNS]). Death occurs in a few hours to days.

Above 0.5 Gy Above 1.0 Gy

Above 3.0 Gy

3.0 to 5.0 Gy

8.0 to 15.0 Gy

Above 10.0 Gy > 25 Gy

18

*Source: AERB lecture notes

Biological Effects of Radiation Early Somatic Effects (Partial Body Irradiation) Partial body exposure to the above dose ranges produces only local effects. A whole body exposure to a dose of 4 Gy can be lethal. But exposure of a part of the body to the dose will not be life threatening. However, it can produce certain serious local effects (Table 2.3). The seriousness of the local effects too depends on the dose rate and the period of exposure etc. All the early somatic effects do have a threshold dose, below which they do not occur. Beyond the threshold dose, severity of the effect increases with the dose. Since the threshold doses are much higher than the normal occupational dose, these effects may not occur from normal occupational exposure to radiation. Table 2.3: Early somatic effects due to acute partial body exposure to low LET radiation Dose

Region

Effect

0.15 Gy 3.5 – 6.0 Gy 1.5 – 2.0 Gy 2.5 – 6.0 Gy 3 Gy 5 Gy 6 Gy

Testes Testes Ovaries Ovaries Hair follicles Eye Skin

10-20 Gy

Skin

Temporary sterility Permanent sterility Temporary sterility Permanent sterility Epilation (fall of hair) Cataract (after 2-3 years) Reddening of skin (erythema) Permanent epilation Burns, blisters, wounds, death of tissues

Late Somatic Effects Exposure to low levels of radiation over a prolonged period, may lead to late effects. Persons who recover from early somatic effects too may develop late somatic effects in life. Late effects are characterized by latent period, which can be as long as 30 years. The important late effects are cataract and cancer. Dose needed to produce cataract may be greater than 8 Gy of fractionated irradiation (not acute) with low LET radiation and the latent period may be 5-10 years. The biological process required to transform the damaged cells to a cancer cell is very complex and the latent period may vary from 2-5 years for leukemia (blood cancer) and 5-30 years or more in the case of cancers of the lung, bone etc. Hereditary Effects Hereditary effects occur in the progeny of exposed individuals when reproductive cells carrying radiation-induced damages (mutations) participate in the process of fertilization. Only that amount of radiation dose to reproductive organs, which occurs up to the time of conception, can affect the general characteristics of the offspring. However, the present 19

Textbook of Radiological Safety knowledge on hereditary effects is limited to laboratory animals. No evidence for increase in hereditary effects was observed in human population exposed to both high and low dose of radiation. Hereditary and both early and late somatic effects of radiation can be classified into two categories, namely (i) deterministic effects and (ii) stochastic effects. Deterministic Effect A deterministic effect (non stochastic) is one “which increases in severity with increasing absorbed dose in affected individuals”. It results in cell killing due to degenerative changes in the exposed tissues. It may appear at higher doses (> 0.5 Gy) and soon after the dose is received. It have threshold dose, below which the effect is not seen. All somatic effects except cancer, as mentioned above are deterministic effects of radiation. All these effects will definitely appear in the exposed individual, if the radiation dose received is above the respective threshold doses. Examples are skin erythema, epilation, organ atrophy, fibrosis, cataract, blood changes, and reduction in sperm count. The radiation dose required to produce such effects are very large and are likely to occur only as the result of radiation accidents and patients irradiated in radiotherapy. It is unlikely in diagnostic Radiology, both for patients and occupational workers. As such deterministic effects can be completely avoided by limiting the dose levels well below the threshold doses. Stochastic Effect A stochastic effect is one in which “the probability of occurrence increases with increasing absorbed dose rather than its severity”. It is very important at very low levels (< 0.5 Gy). Any dose, however small, is effective for a certain level of risk for induction of stochastic effects. The risk increases as the dose increases. It have no threshold dose and the chance of occurrence increases with dose and independent of sex and age. Stochastic effects are the principle health risk from low level radiation, which is likely in diagnostic radiology and Nuclear medicine. Radiation induced cancer and genetic effects are examples for stochastic effects. Hence the risk of stochastic effects cannot be completely avoided. However, it can be minimized to an acceptable level. Radiation Induced Cancer Radiation induced cancers may appear only some years after the radiation was received. This is known as latent period, which may extend from 10 to 30 years. The risk of inducing a particular type of cancer is measured by comparing the number of cancers produced in the irradiated population sample in excess of those expected in the same size of un irradiated population sample. This is known as Risk factor for the particular cancer. 20 Leukemia has the highest risk factor among all cancers, since tissue cells

Biological Effects of Radiation are rapidly dividing. There is no safe dose limit and all doses of radiation carry some form of risk. Hence, radiation induced cancers can not be prevented, only be reduced by minimizing the radiation dose. Genetic Effects The radiation effects produced in the successive generation of the exposed individual are called genetic effects. Genetic effects are caused by radiation induced damage to the genes or chromosomes in the ova or spermatozoa. The above damage is due to ionization and subsequent faulty recombination of the molecules which make up the chromosomes. As a result the biological code contained by the genes on the chromosomes gets altered and produce structural abnormality in the chromosome. It will be passed on to the future generations if reproduction takes place. Severe genetic effects are not observed due to short life span of the individual and inability to reproduce. Only less severe effects are seen in the human population as a result of reproduction. Epidemiological studies carried out on 30,000 children born to the survivors of atom bomb (500 mSv) do not show an increase in the incidence of genetic disorders. There is no direct evidence of either elevated cancer risks or genetic disorders among human population exposed to low level radiation. To protect the future generations, minimize the use of artificial radiations and adopt protection devices such as gonad shield in children, during radiography. Fig. 2.3, summarizes both early and late biologic effects due to radiation exposure from 0.001 mSv to 20 Sv.

Fig. 2.3: The early and late biologic effects of radiation with dose ranging from 0.001 mSv to 20 SV

21

Textbook of Radiological Safety RADIATION EFFECTS ON DNA The ionizing radiation deposits energy in the DNA molecule and produce chemical changes, which leads to structural changes. This includes (i) hydrogen bond breakage, (ii)molecular dehydration, and (iii) intermolecular and intramolecular cross linking. The rupture of hydrogen bond, which link the DNA base pairs, may result in irreversible structural changes in the molecule. DNA molecular breakage may occur as single strand breaks, double strand breaks, base loss or base changes occurring in DNA. Single strand breakage between the sugar and the phosphate can rejoin. Oxygen can cause the broken strand become peroxidized, and prevent it from rejoining. Thus presence of oxygen potentiates the radiation damage. A double strand break are genotoxic lesions that can result in chromosome aberrations. This may lead to carcinogenesis through activation of oncogenes, inactivation of tumor suppressor genes, or loss of herterozygosity. Single strand breaks are easily repairable than double strand breaks as far as the low LET radiation is concerned (Fig. 2.4).

Fig. 2.4: Radiation action on DNA molecule: Single and double strand breaks

Molecular cross linking is another common structural change, as result of reactive sites at the point of chain breakage. Cross linking may be between two DNA molecules, DNA and a protein, two base pairs within DNA. The 22 loss of change of base is considered as mutation.

Biological Effects of Radiation Chromosome breaks produced by radiation do occur and can be observed microscopically. Chromosomal damage that occurs before DNA replication is called chromosome aberrations, whereas that occur after DNA synthesis is called chromatid aberrations. In the later, only daughter of the one of the damaged chromatids pair is affected. Repair of chromosomal aberrations depends on stage of the cell cycle, and type and location of the lesion. There is a strong force of cohesion between broken ends, resulting in rings and dicentrics formation. Chromosomal aberrations in human lymphocytes can be scored and used as a biological dosimeter to estimate radiation dose. For this lymphocytes are cultured from human blood sample. It is stimulated to divide, allowing karyotype to be performed. The cells are arrested at metaphase, and the frequency of rings and dicentrics are scored. Total body doses > 0.25 Gy can be estimated in the above method. RADIATION EFFECTS IN UTERO Developing embryo is extremely sensitive to ionizing radiation, as it is characterized by cell proliferation, migration and differentiation. As per the Bergonie and Tribondeau’s laws of radiosensitivity, highly differentiated cells are more radiosensitive. The response after radiation depends upon the total dose, dose rate, radiation quality and gestation period of the pregnant women. The type of damage includes prenatal death, congenital abnormalities, growth impairment, reduced intelligence, genetic aberrations and an increased cancer risk. Gestational periods is divided into three stages namely (i) Preimplantation, (ii)major organogenesis, and (iii) fetal stage. These stages vary with radiosensitivity and result in different radiation response at each stage. Preimplantation begins with the union of the sperm and egg and continue until the zygote is embedded in the uterine wall. During this period (9-14 days) the conceptus is more sensitive, any radiation damage may result in prenatal death. The congenital abnormalities are low, though it is not completely absent. The most sensitive time of exposure in humans is 12 hr after conception (when two pronuclei fuse together to one cell) and again at 30 and 60 hr, when the first two divisions occurs. Animal study revealed an increase in the spontaneous abortion rate after a dose of about 50-100 mGy during the preimplantation period. Later a radiation dose of >250 mGy is required induce prenatal death and the spontaneous abortion rate is reported as 30-50%. Major organogenesis is the period between 2-8 wk (15-50 days) after conception. It is the critical period for radiation induced birth defects in human. The embryonic malformations includes cataract, mental retardation, microcephaly, microphthalmia, growth retardation, skeletal defects and 23 carcinogenesis. Exposure >100 mGy from the Hiroshima atomic bomb

Textbook of Radiological Safety resulted an increase in the incidence of microcephaly. Exposures in excess of 1Gy are associated with a high incidence of CNS abnormalities. The fetus stage is between 50-280 days, prenatal death and congenital anomalies are negligible. However growth retardation, abnormalities of the nervous system and sense organs are the primary. There is risk of childhood leukemia is significant. The damage manifests later in life as behavioral alterations or reduced intelligence (IQ). RADIATION RISK Radiation risk is a probability, that a given individual will incur a deleterious effects as a result of a dose of radiation. It includes (i) somatic risk (ii) genetic risk and (iii) fetal risk. Scientists have developed dose response models to predict the risk of cancer in human populations, due to low level radiation exposure. These models led to dose response curves, whose shape are non threshold linear, linear quadratic and quadratic as shown in Fig. 2.5.

Fig. 2.5: The linear and non linear models of radiation effects

24

A non threshold linear curve overestimates the incidence of cancer at lower doses from low LET radiation. Hence, it is used in radiation protection for estimating risk. A linear quadratic dose response model predicts lower incidence of cancer than the linear model at low doses and higher incidence at intermediate doses. The risk estimates were revised in 1990 after reassessing the Japanese atomic bomb survivors, which resulted an increase in risk estimation. The reasons for the above are (i) revised neutron dosimetry which incorporated greater risk from low LET radiation, (ii) large number of cancer cases is seen among bomb survivors, and (iii) revised model projects risk beyond the period of observation.

Biological Effects of Radiation Risk Models There are two type of risk models namely (i) Mutiplicative, and (ii) Additive risk models (Fig. 2.6). The multiplicative model predicts that, after a latent period, the excess risk is a multiple of the natural age specific cancer risk for a given population. It accounts for age and cancer risk at the time of exposure. The risk increases with age (predicts greater risk at older age). The Fig. 2.6(a) describes the differences in magnitude of the projected cancer risk for a given exposure at different ages. The additive risk model which predicts a fixed increase in risk unrelated to the spontaneous age-specific cancer risk at the time of exposure (Fig. 2.6(b)). In this model a constant increment in incidence is added to the spontaneous disease incidence through out the life. Both models do not describe cancer risk adequately, after exposure. The necessary modification in the risk estimate is provided by the BEIR report, which is explained in later paragraphs.

(a)

(b)

Fig. 2.6: The multiplicative model and additive model

Relative Risk Relative risk (RR) is another way of expressing the risk from radiation in an exposed population. It is the ratio of the cancer incidence in the exposed population to that in general population. For example, a relative risk of 1.2, predicts a 20 % (120/100) increase over the spontaneous rate of cancer incidence. The excess relative risk = RR-1= 1.2-1=0.2 To detect the above relative risk (1.2), with a statistical confidence of 95% (p< 0.5),when the spontaneous incidence in the population is 2 %, one must require a study sample of population of >10000.

25

Textbook of Radiological Safety Absolute Risk Another way of expressing risk is the absolute risk. It is expressed as number of excess radiation induced cancers per 104 people /Sv-year. Example 1: The risk is 4 per 10,000 person - Sv and the latency period is 10 years. What is the risk of developing cancer in the next 40 years, from a dose of 0.1 Sv? Actual duration = 40 – 10 = 30 years Dose = 0.1 Sv Risk = 4 × 10-4 Risk of developing cancer = 30 × 0.1 × 4 × 10-4 = 12 × 10-4 If 10000 people are exposed to a dose of 0.1 Sv, 12 additional cases of cancer will be seen in that population in the next 40 years. BEIR Report V and VII Risk Estimate

26

The National research council committee on the Biological effects of ionizing radiation (BEIR) report V has been published in 1990 on the title “Health effects of exposure to low levels of ionizing radiation”. As per the report the radiation induced mortality at low exposure level is 4 % per Sv and it is 8% for high dose rates. This is in agreement with the ICRP estimate, which is 4 % per Sv for a population of adult workers and 5 % per Sv for the whole population (includes young ones). The BIER V committee advocate the linear dose response model for all cancers except leukemia and bone cancer, to which linear quadratic model is recommended. BEIR VII report (2003) gives the most up-to-date and comprehensive risk estimates for cancer and other health effects from exposure to low level ionizing radiation. That is, it accounts for both cancer incidence and cancer mortality. It supports a “linear-no-threshold” (LNT) risk model—that the risk of cancer proceeds in a linear fashion at lower doses without a threshold and that the smallest dose has the potential to cause a small increase in risk to humans. There is a linear dose-response relationship between exposure to ionizing radiation and the development of solid cancers in humans. It is unlikely that there is a threshold below which cancers are not induced, but at low doses the number of radiation-induced cancers will be small. Other health effects (such as heart disease and stroke) occur at higher radiation doses, but additional data must be gathered before an assessment of any possible dose response, that can be made between low doses of radiation and non cancer health effects. The report also concludes that with low dose or chronic exposures to low LET irradiation, the risk of adverse heritable health effects to children conceived after their parents have been exposed is very small compared to baseline frequencies of genetic diseases in the population.

Biological Effects of Radiation Radiation related cancer mortality risks for woman averaged is 37.5% higher than for men in the solid tumors. Exposure in infants, as compared to adults, produces 3-4 times the cancer risk. Female infants have almost double the risk of males. BEIR VII lifetime risk model predicts that approximately one individual in 100 persons would be expected to develop cancer (solid cancer or leukemia) from a dose of 100 mSv while approximately 42 of the 100 individuals would be expected to develop solid cancer or leukemia from other causes. Lower doses would produce proportionally lower risks. For example, it is predicted that approximately one individual in 1000 would develop cancer from an exposure to 10 mSv. Somatic Risk Somatic risks may arise from both stochastic and deterministic effects. Cancer induction is the largest risk of radiation. Bone marrow, gastrointestinal tract mucosa, breast tissue, gonads and lymphatic tissue are most susceptible for cancer induction. Caner risk are higher for children than for adults. Radiation may induce both benign and malignant tumors with latent period. Cancer risk is estimated as 4 × 10-2 /Sv. The ICRP recommended risk factors are given in the Table 2.4 below. Table 2.4: ICRP 60 Risk factors Assumed radiation risks

ICRP Publication 60

Workers

4.0 × 10-2 Sv-1 for fatal cancer 0.8 × 10-2 Sv-1 for non fatal cancer detriment 0.8 × 10-2 Sv-1 for severe genetic effects 5.0 × 10-2 Sv-1 for fatal cancer 1.0 × 10-2 Sv-1 for non fatal cancer 1.3 × 10-2 Sv-1 for severe genetic effects Not specifically stated

Members of the public Embryo-fetus

Genetic Risk Genetic risk is the result of radiation exposure to the gonads. Genetic risk analysis assume that the (i) exposed population consists of all ages and both sexes and (ii) severe genetic effects in the next two generations. The genetically significant dose (GSD) is an index to estimate the radiation induced mutation in germ cells of a given population. The sensitivity of a population to radiation induced damage is measured by doubling dose. It is defined as the dose required per generation to double spontaneous mutation rate. The spontaneous mutation rate is about 5 × 10-6 per locus and 15 × 10-4 per gamete for chromosome abnormalities. The doubling dose for humans is estimated as ~1Gy per generation, which 27 is extrapolated from animal data.

Textbook of Radiological Safety A dose 100 mGy would produce only about 200 additional genetic disorders per 1 million live births (0.02 % per 100 mGy) in the first generation, whereas the normal incidence is about 1 in 20 (5%). Hence the 100 mGy dose can cause additional genetic disorders of about 0.4 % {(5.02-5)/5) × 100} only. The usual diagnostic and occupational exposures would not be expected to result in any significant genetic risk to their progeny. Fetus Risk Doses lower than 100 mGy generally carry negligible risk. To avoid congenital abnormalities abortion may be advised, only when doses are >100 mGy. The fetal dose from Medical Diagnostic procedures rarely exceeds 50 mGy. This will not put the fetus any significant increase in risk for congenital malformation or growth retardation. It is estimated that the excess risk of childhood cancer from in utero irradiation is approximately 6% per Gy. The relative incidence of various health effects with radiation exposure in utero at various stages of gestation are shown in Fig. 2.7.

Fig. 2.7: Health effects of radiation exposure in utero at various stages of gestation

Radiopharmaceuticals administered to pregnant women is associated with radiation risk to the fetus. There are two type of radiopharmaceuticals: (i) those that cross placenta and (ii) those that remain on the maternal side. For example radioiodine can cross the placenta and concentrate on the fetal thyroid after 13 th week of gestation. It may be 55 to 75 % between 14 and 28 22 weeks of gestation. This may result in hyperthyroidism or ablation.

Biological Effects of Radiation Estimate shows that the dose to the fetal thyroid may range from 230 to 580 mGy / MBq for gestational period of 3 and 5 months. Total body fetal dose estimate ranges from 0.072 to 0.27 mGy / MBq during pregnancy. Example 2: Calculate the risk for radiation workers and the public for a annual effective dose limit of 20 mSv and I mSv respectively (Assume the risk coefficient is 4 × 10-2 per Sv). Worker: Annual dose limit = 20 mSv = 0.02 Sv Annual risk = 0.02 × 4 × 10-2 = 8 × 10-4 Public: Annual dose limit = 1 mSv = 0.001 Sv Annual risk = 0.001 × 4 × 10-2 = 4 × 10-5 TEN DAY RULE AND ITS PRESENT STATUS ‘Ten day rule’ was postulated by ICRP for woman of reproductive age. It states that “whenever possible, one should confine the radiological examination of the lower abdomen and pelvis to the 10 day interval following the onset of menstruation.” The original proposal was for 14 days, but this was reduced to 10 days to account for the variability of the human menstrual cycle. In most situations, there is growing evidence that a strict adherence to the “ten day rule” may be unnecessarily restrictive. When the number of cells in the conceptus is small and their nature is not yet specialized, the effect of damage to these cells is most likely to take the form of failure to implant, or of an undetectable death of the conceptus; malformations are unlikely or very rare. Since organogenesis starts 3 to 5 weeks post conception, it was felt that radiation exposure in early pregnancy couldn’t result in malformation. The main risk is that of abortion if the radiation exposure results in death of the conceptus. It requires a fetal dose of more than 100 mGy for this to occur. Based on this, it was suggested to do away with the 10 day rule and replace it with a 28 day rule. This means that radiological examination, if so justified, can be carried throughout the cycle until a period is missed. Thus, the focus is shifted to a missed period and the possibility of pregnancy. If there is a missed period, a female should be considered pregnant unless proved otherwise. In such a situation, every care should be taken to explore other methods of getting needed information by using non radiological examinations. BIBLIOGRAPHY 1. AERB short course on radiation safety, lecture notes 2008. 2. BEIR VII: Health Risks from Exposure to Low Levels of Ionizing Radiation, National Academies Press, 500 Fifth Street, NW, Washington, DC 20001; 800: 624-6242; www.nap.edu. 3. Donald TG, Paul C, Martin Vosper. Principles of Radiological physics, (5th edn.) Churchill Livingstone 2007.

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Textbook of Radiological Safety 4. Jerrold TB, Seiber JA, Edwin ML, John MB. The essential physics of medical imaging, (2nd edn.), Lippincott Williams & Wilkins 2002. 5. Mettler FA, Upton AC. Medical effects of ionizing radiation, (2nd edn.), Philadelpia:WB saunders, co. 1995. 6. National research council, Committee on the biological effects of ionizing radiations. Health effects of exposure to low levels of ionizing radiation (BEIR V). Washington, DC: National Academy Press, 1990.

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Chapter

3

Radiation Exposure Control

The four principal methods by which radiation exposures to persons can be minimized are (i) Time (ii) Distance, (iii) Shielding, and (iv) Contamination control. The following paragraphs will explain their role and application in medicine. TIME The total dose received by a radiation worker is directly proportional to the total time spent in handling the radiation source. Lesser the time spent near the radiation source, lesser will be the radiation dose. As the time spent in the radiation field increases, the radiation dose received also increases. Hence, minimize the time spent in any radiation area. Techniques to minimize time in a radiation field should be recognized or practiced. All radiation sources do not produce constant exposure rates. Diagnostic X-ray machines typically produce high exposure rates over brief time intervals. For example, chest X-ray produces an skin entrance exposure of 20 mR in less than 1/20 of a second, equivalent to 1440 R/hr. Hence, radiation exposure can be minimized by not energizing the X-ray tube, when personnel are nearer to the machine. Nuclear medicine procedure produce lower exposure rate for extended periods of time. The exposure rate at 1m from a patient injected with 20 mCi of Tc-99m, for bone scan is 1 mR/hr. It reduces to 0.5 mR/hr after 2 hours, due to decay and urinary excretion. Hence both the knowledge of exposure rate and how it changes with time are important elements in reducing personnel exposures. The time spent near the radiation source can be minimized by under standing the task to be performed and the suitable equipment to complete them in short interval with safety. Hence, one has to plan the radiation procedure, practice the procedure with out radiation and share the essential duties, to reduce radiation exposure. For example, fluoroscopy screening time should be kept short by the use of last frame hold facility, in addition to the use of foot switch. Example 1: A radiographer is performing barium examination under fluoroscopy and the equipment is ‘ON’ for 3 minutes for each examination. The radiation level at the location of the radiographer is 100 mR/h. How many such procedures the radiographer can carry out per week?

Textbook of Radiological Safety The annual equivalent dose limit prescribed for the radiographer is (occupational worker) 20 mSv = 2000 mrem ≈ 2000 mR 2000 mR = 40 mR 50 weeks Exposure rate at the location of radiographer = 100 mR/h

The permitted weekly dose =

=

100 mR/min 60

100 mR × 3 min = 5 mR 60 min Hence, the number of procedures the radiographer can associate with

The exposure in each procedure =

in one week =

40 mR =8 5 mR

Example 2: An operator is handling 10 mCi of I-131 source with 30 cm tongs. Within how much time the technician will receive the weekly permissible equivalent dose? (assume 1R=1 rad, Γ20 =2.18 R-cm2/mCi-hr for I-131) Exposure level at 30 cm from 10 mCi of I-131 source = 2.18 × 10 mCi /(302) = 0.024 R/hr = 24 mR/hr Weekly permissible exposure Allowed time of work =

2000 mR = 40 mR 50 weeks

40 mR = 100 minutes 24 mR/hr

DISTANCE Radiation intensity (exposure rate) from a point source decreases with distance, due to divergence of the beam. It is governed by the inverse square law, which states that the exposure rate from a point source of radiation is inversely proportional to the square of the distance. If the exposure rate is X1 at distance D1, then the exposure rate X2 at another distance D2 is given by ⎛D ⎞ X 2 = X1 ⎜ 1 ⎟ ⎝ D2 ⎠

2

(1)

Doubling the distance from the X-ray source decreases the X-ray beam intensity by a factor of 4. If the exposure rate is 100 mR/hr at 1 m, then it will be 25 mR/hr at 2 m (Fig. 3.1). Larger the distance, lesser will be the 32 radiation dose. This relationship is valid for point sources only, whose

Radiation Exposure Control dimensions are very small compared to distance under consideration. Thus the relationship is not valid near (<1 m) a patient injected with radioisotopes, since the exposure rate decreases less rapidly than inverse square law.

Fig. 3.1: The inverse square law: As the distance increases by a factor 2, the radiation intensity decreases by a factor of 4

In diagnostic radiology at 1m from a patient, the scattered radiation is about 0.1-0.15% of the intensity of the primary beam. Hence, all personnel should stand as far away as possible during X-ray procedures. Personnel should stand at least 2 m from the X-ray tube and the patient and behind the shielded barrier or out of the room, whenever possible. In nuclear medicine it is difficult to shield the technologist from the radiation emitted from the patient, during patient imaging and distance is the only primary dose reduction method. The nuclear medicine technologists receive the annual dose mainly by patient imaging. Hence, imaging rooms should be designed to maximize the distance between the source and control console. Because most of the time the technologists spend their time at the control console for acquisition and image processing. Unshielded radiation sources should never be manipulated by hand. Tongs or other handling devices are used to increase the distance between source and hand. A typical exposure rate with distances, for a nuclear medicine facility is given in Table 3.1. Table 3.1: Exposure rate with distance in nuclear medicine Radionuclide10 mCi Ga-67 Tc-99m I-123 I-131 Xe-133 Tl-201

Exp rate (mR/hr) at 1cm 7500 6200 16300 21800 5300 4500

Exp rate (mR/hr)at 10 cm 75 62 163 218 53 45

Example 3: The exposure rate from a fluoroscopic X-ray machine is 5 R/ min at 50 cm. What would be the exposure rates at (i) 40 cm, and (ii) 60 cm?

33

Textbook of Radiological Safety 1. X1 = 5 R/min, D1 = 50 cm, D2 = 40 cm , X2 = ?

⎡ X 1 × ( D1 )2 ⎤ ⎡ 5 R / min× ( 50 cm )2 ⎤ ⎣ ⎦ ⎣ ⎦ X2 = = = 7.81 R/ min 2 2 ( 40 cm ) ( D2 ) 2. X1 = 5 R/min, D1 = 50 cm, D2 = 60 cm , X2 = ? ⎡ 5 R / min× ( 50 cm )2 ⎤ ⎣ ⎦ X2 = = 3.47 R/ min 2 ( 60 cm )

Example 4: What would be the radiation level at 10 cm from a 5 mCi source of Co-60? (Given, Γ = 13 R-cm2/mCi-hr for Co-60) Exposure level at 1 cm from 5 mCi Co-60 source = 13 × 5= 65 R/hr X1 = 65 R/hr, D1 = 1 cm, D2 = 10 cm , X2 = ?

⎡ X 1 × ( D1 )2 ⎤ ⎡ 65 R / hr × ( 1 cm )2 ⎤ ⎣ ⎦ ⎣ ⎦ = = 0.65 R/ hr = 650 mR/hr X2 = 2 2 10 cm D ( ) ( 2) Example 5: What would be the distance required to reduce the radiation level from a 10 mCi Ir-192 source to 2mR/hr? (Given, Γ=4.69 R-cm2/mCihr for Ir-192) Exposure rate at 1cm from 10 mCi Ir-192 source = 4.69 × 10 = 46.9 R/hr = 46900 mR/hr X1 = 46900 mR/hr, D1=1 cm, D2= ? cm, X2 = 2 mR/hr ?

⎡ X 1 × ( D1 )2 ⎤ ⎣ ⎦ X2 = 2 ( D2 ) ⎡ 46900 mR / hr × ( 1 cm )2 ⎤ ⎦ , D2 = 153 cm or 1.53 m 2 mR/hr = ⎣ 2 ( D 2 cm ) SHIELDING When maximum distance and minimum time do not ensure an acceptably low radiation dose, adequate shielding must be provided, so that radiation beam will be sufficiently attenuated. The material that attenuates the radiation exponentially is called a shield and the shield will reduce exposure to patients, staff and the public. When a photon passes through an absorber (Fig. 3.2) of thickness x, both absorption and scattering takes place. As a result, the transmitted beam will have less number of photons. This can be represented by the relation 34 I = I0 e- μx (2)

Radiation Exposure Control where I is the number of transmitted photons, I0 is the number of incident photons, e is the base of natural logarithm and μ is the linear attenuation coefficient.

Fig. 3.2: Attenuation of radiation through a shielding material

HALF VALUE LAYER The linear attenuation coefficient is defined as the reduction in the radiation intensity per unit path length and is expressed in cm-1. The linear attenuation coefficient is related to the term half value layer (HVL) as follows:

μ=

0.693 HVL

(3)

The thickness of material that attenuates radiation beam by 50 % is the half-value layer or half value thickness (HVT). The reduction factor offered by one HVT is 2 and by 2 HVT is 2 × 2 or 22. Hence, the reduction factor offered by n HVT of the shielding material is 2n. For diagnostic X-ray beam energies, the HVL for soft tissue ranges from 2.5 to 3.0 cm. The Table 3.2 indicates the decrease of radiation intensity with increasing HVT’s (for heavily filtered X-ray beam).

35

Textbook of Radiological Safety Table 3.2: Relation between half value layer and % of transmission No. of HVTs

% transmission

0 1 2 3 4 5 6 10

100.0 50.0 25.0 12.50 6.25 3.12 1.56 0.09

Tenth Value Layer There is another interesting term called tenth value layer (TVL or TVT), which gives the thickness of material that attenuates the radiation beam by 90 %. The reduction factor offered by one TVT is 10, and by 2 TVT is 10 × 10 or 102 and 10n for n TVT and so on. The TVT and HVT can be related by using the equation: TVT = =

2.303 μ

(4)

2.303

( 0.693 / HVT )

= 3.32 HVT For a given material, both HVT and TVT depend upon the energy of the incident radiation and the shielding material. Table 3.3 shows the HVT and TVT values for gamma rays with respect to concrete, steel and lead. Table 3.3: HVT and TVT values of concrete, steel and lead for Ir-192,Cs-137 and Co-60 (IAEA Safety series 47) Source IR-192 CS-137 Co-60

Concrete, mm HVT TVT 43 48 62

152 175 218

Steel, mm HVT TVT 13 16 21

43 53 71

Lead, mm HVT TVT 6 6.5 12

16 22 41

The shielding type, thickness and the location are functions of photon energy, number, intensity, source geometry and exposure rate. Hence, the reduction in intensity depends upon the nature and thickness of the shield and energy of the radiation. The thicker the shielding, lesser the radiation. In diagnostic energy range, X-ray attenuation is primarily by photoelectric 36 effect and to some extent by compton effect. Photoelectric effect varies with

Radiation Exposure Control atomic number (Z) and is proportional to Z 3. Hence, one can compare lead (Z=82) and aluminium (Z=13) as shielding material for the same thickness (g/cm2), with their atomic numbers. The reduction in intensity caused by lead over aluminium is given by 3

⎛ 82 ⎞ ⎜ ⎟ = 250.3 ⎝ 13 ⎠ In other words, 1 gm/cm2 of lead will be as effective as 250 g/cm2 of aluminium. Hence in many situations, lead is used as shielding material to reduce the volume of shielding. Lead, brick and concrete, are the most commonly used materials for shielding radiation exposure in medicine. Lead aprons, gonad shield, lead gloves, lead bricks, and lead glass are also used as shield. The NCRP-49 (1976) describes the structural shielding design and evaluation for Medical use of X-rays and γ rays of energies up to 10 MeV. When the report was made single phase generator was in use, the annual dose limit is 100 mrem per year, neglected the image receptor attenuation, and considered low speed film-combination. Hence, strict adherence of NCRP-49 would result in expensive shielding procedures. The factors to be considered while determining the amount and type of shielding are (i) ALARA principle, personnel radiation must be kept As Low As Reasonably Achievable, (ii) Personnel exposures should not exceed the regulatory limits. This is 30 mSv for occupational workers and I mSv for the general public in India. The corresponding limits in USA is 50 mSV and 1 mSv respectively and (iii) The dose equivalent to the controlled area (restricted entry) should not exceed 0.6 mSv per week. The corresponding limit for the uncontrolled area is 0.02 mSv per week.

SOURCES OF EXPOSURE The radiations used in medical application appear in three forms namely (i) primary radiation, (ii) scattered radiation and (iii) leakage radiation. The scattered and leakage radiations are jointly called as secondary radiation (Fig. 3.3). Primary Radiation The radiation passing through the open area defined by the collimator of the X-ray tube / radiation equipment is called primary radiation. The amount of primary radiation depends on output, number of patients per week, and fraction of time the radiation beam is directed towards any particular location. Concrete, lead and steel are used as primary barrier shielding material, depending upon the structural and spatial considerations. For example concrete is preferred for the construction of walls and ceiling, because it is cheep. Whenever there is shortage of space, 37 lead and steel can be used as primary barriers.

Textbook of Radiological Safety

Fig. 3.3: The various sources of exposure namely primary, scattered and leakage radiations

Scattered Radiation Scattered radiation from a patient, tabletop, collimator or shield is also a source of radiation. Scattered radiation is due to the interaction of the primary with the patient, causing a portion of the primary radiation to get redirected. The fluence of scattered radiation depends on (i) the volume of the patient irradiated, (ii) spectrum of the primary beam, (iii) field size, and (iv) scattering angle. Scattered radiation must be considered as a separate source, for radiation safety purposes and in general it has low energy. In diagnostic radiology, the quality of the scattered radiation is assumed to be same as that of the incident beam. It is about 0.1-0.15% of primary at 1m from patient, for a field area of 20 × 20 sq cm. In mega voltage radiations, the ratio of scattered dose to the incident dose is denoted as α, which varies with scatter angle and beam quality. The α is assumed to be 0.1% of the primary for 90 degree scatter. The maximum energy of the 90° scattered photon is 500 keV. The transmission of the above beam is similar to that a 500 kVP primary beam. The penetrating power of scattered beam is greater at smaller scatter angle. Also greater fraction of primary is scattered at smaller angles. Leakage Radiation Leakage radiation is the one that emanates from the source housing, other

38 than primary. The quality of the leakage radiation is approximately the

Radiation Exposure Control same as that of the primary beam. Since it passes through the source housing, its effective energy is very high in diagnostic X-rays. LEAKAGE LIMITS Leakage Limits for X-ray Housing The leakage radiation through the protective X-ray tube housing in any direction, averaged over an area not larger than 100 cm2 with no linear dimension greater than 20 cm, shall not exceed an air kerma of 1 mGy in one hour at a distance of 1.0 m from the X-ray target. This will be applicable to all types of X-ray tubes except mammography. In mammography the corresponding leakage limit is 0.02 mGy at 5 cm from any point on the external surface of X-ray tube housing (AERB /SC/MED-2). The NCRP report 102 supersedes the previous NCRP report 33 and recommend the following limits: 1. 5-50 kVp: The leakage exposure rate shall not exceed 0.1 R in any 1 hour at any point 5 cm from the source assembly. 2. 50-500 kVp: The leakage exposure rate at a distance of 1 m from the source shall not exceed 1 R in any 1 hour. In addition, these assemblies shall limit exposure rate to 30 R/h at 5 cm from the surface of the assembly. 3. Greater than 500 kVp: The absorbed dose rate due to leakage radiation (excluding neutrons) at any point outside the maximum field size, but within a circular plane of radius 2 m that is perpendicular to and centered on the central axis at the normal treatment distance, shall not exceed 0.2% of the useful beam dose rate at the treatment distance. Except for the area defined above, the leakage dose rate from the source assembly at any point at a distance of 1 m from the electron path between the source and the target shall not exceed 0.5% of the useful beam dose rate at the treatment distance. The neutron contribution to the dose within the useful beam shall be kept well below 1% of the X-ray dose. Outside the useful beam, the neutron dose should be reduced to as low as practicable. Leakage Limits for Cobalt Teletherapy For Cobalt teletherapy, in the beam off position, the leakage radiation measured at a distance of 1 m from the source shall not exceed 20 μGy/h. At any readily accessible position 5 cm from the surface of the housing the leakage radiation shall not exceed 200 μGy/hr. In the beam on position, the leakage radiation measured at a distance of 1m from the source shall not exceed either 10 mGy/h or 0.1% of the useful beam air kerma rate at 1 m from the source, whichever is greater (AERB/SC/MED-1). However, the recent NCRP report 102 summarizes the cobalt teletherapy leakage as follows: The leakage dose rate from the source housing with the

39

Textbook of Radiological Safety beam in the OFF position shall not exceed 2 mrad/h on the average and 10 mrad/h maximum in any direction, at a distance of 1 m from the source. With the beam in the ON position, the leakage dose rate from the source housing shall not exceed 0.1% of the useful beam dose rate, both measured at a distance of 1 m from the source. In addition, for sources that give rise to a useful beam dose rate of less than 10,000 rad /h at 1m, the leakage from the source housing shall not exceed 1 rad/h at 1m from the source. Leakage Limits for Brachytherapy The Brachytherapy source storages (LDR/HDR equipments), emergency storages and in house transport containers, must meet the following limits for leakage radiation. 1. The air kerma rate of the leakage radiation measured at any readily accessible position 5 cm from the surface of the storage must not exceed 200 μGy/ hr, and 2. At 1m from the centre of the storage the air kerma rate of leakage radiation must not exceed 20 μGy/hr. 3. In the case of emergency storage container or an in house transport container, the air kerma rate of the leakage radiation measured at a distance of 5 cm from the surface of the container must not exceed 2 mGy/hr and at a distance of 1 m from the source of the container must not exceed 100 μGy/hr (AERB/SC/MED-3). PROTECTIVE BARRIER DESIGN Protective barriers are designed to ensure that the dose equivalent received by any individual does not exceed the applicable maximum permissible value. Barrier should protect all the three types of radiation namely primary, scattered and leakage radiations. A barrier sufficient to attenuate the useful beam to the required degree is called the primary barrier. The required barrier against leakage and scattered (stray) radiation is called the secondary barrier. The following materials are used as barriers: Lead (Most commonly used material because of its high attenuation properties and low cost), concrete, leaded glass, leaded acrylic and gypsum. The amount of attenuation necessary depends on five factors: • Workload • Use factor • Occupancy factor • Distance • Radiation exposure level. Work Load (W) Workload is a measure of expected exposure levels obtained from the patient

40 load and machine ON time. It is expressed in mA-min/week in diagnostic

Radiation Exposure Control radiology (< 500 kVp), which can be obtained by multiplying the maximum mA with beam on time in minutes/week. For mega voltage machines, it is expressed in weekly dose (Gy/week) delivered at 1m from the source. This can be obtained by multiplying the number of patients treated per week with dose delivered per patient at 1m. Use Factor (U) It is the fraction of the operating time during which the radiation under consideration is directed towards a particular barrier (wall) per week. The use factor for different barriers are as follows: • Primary barrier: U=0 to 1 • Secondary barrier: U=1 • Floor: U=1 • Walls: U=1/4 • Ceiling: U=1/4 to 1/2 Occupancy Factor (T) The occupancy factor relates to the amount of time rooms adjacent to the treatment room are occupied. An area below ground would have no occupancy at all and therefore T would equal zero. Areas that are intermittently occupied, such as corridors, would have a slightly greater occupancy and an area such as an office even greater. It is defined as the fraction of the operating time during which the area of interest is occupied by the individual. Table 3.4 shows type of occupancy and occupancy factors (NCRP 49). Table 3.4: Occupancy factors (NCRP 49) Type of occupancy

Occupancy factor (T)

Office, reception, shops, children play areas, staff rooms, control room

1 (Full occupancy)

Corridor

1/4 (Partial occupancy)

Toilet, bath rooms, outside areas with seating, store rooms

1/16 (Occasional occupancy)

Distance (d) It is the distance in meters from the radiation source to the area to be protected. Inverse square law is assumed for both primary and stray radiation. Radiation Exposure Level (XT or P) It represents the exposure (mR/week) at a given location in an adjacent area. It is function of technique, workload, primary, scattered and leakage 41 level and corresponding distance to the point of interest.

Textbook of Radiological Safety FACILITY DESIGN FOR DIAGNOSTIC X-RAYS Workload A busy diagnostic X-ray unit will have a workload of 1,000 mA.min/week, whereas small clinics may have 100 mA.min/week. For example, the workload of a hospital with 20 patients per day, 3 films per patient, and 50 mAs per film will be calculated as follows: W = (20 patients/day) × (5 days/week) × (3 films/patient) × (5 mAs/film) × (1 min/60 s) = 25 mA-min/week In general higher kVp settings decrease the workload. This is due to (i) increase of output (mR/mAs) as the kVp increases and (ii) less attenuation of the incident beam by the patient reduces the mAs, at higher kVp. The workload values for various types of radiographic rooms are given in the Table 3.5 (NCRP report No.49). These are overestimated values as it is based on slower (speed100) film-screen receptors. Table 3.5: Workload for diagnostic X-ray Procedure

Patient load/day

W (mA. min/wk), 100 kVp

W (mA. min/wk), 125 kVp

60 24 24 8

250 750 1000 700

150 300 400 280

Chest,3 films/patient Fluoroscopy General radiography Special procedures

Shielding Calculation for Diagnostic X-ray The X-ray tube output (mR/mAs) is determined by direct measurement. Otherwise it can be obtained from the graph between mR/mAs verses X-ray tube potential as a function of HVL (Fig. 3.4). The primary beam exposure per week (XP) at 1m from the source, is the product of W and the tube current measured at 1m from the source. For example an X-ray tube operating with a workload of 300 mA min/week with out put is 5 mR/mAs at 1m, then XP = workload × tube out put

(5)

= 300 (mA.min/wk) × 5 (mR/mAs) = 300 (mA.min/wk) × 60 × 5 (mR/mA.min) = 90,000 mR/wk The exposure due to scattered radiation from the patient is a fraction of the incident primary exposure. The scatter fraction (S) is 0.15% at 1m for a 42 400 cm2 field size at 125 kVp. This will vary with field size, kVp and the

Radiation Exposure Control

Fig. 3.4: The Tube output in mR/mAs at a distance of 1 m with variation of HVL in mm for Al. (Inverter generator,5% ripple,1.2 mm Al inherent filtration)

scattering angle. The scattered radiation (Xs) is the product of the incident exposure, scatter fraction, and the ratio of the maximum field size relative to 400 cm2. Hence, the scattered radiation exposure per week at 1m from the scatter is ⎛ Xp XS = ⎜ 2 ⎝ dsca

S

⎞ field size ⎟×S× 400 ⎠

(6)

= scatter fraction (0.15%)

dsca = distance to the scatter For example, a chest imaging is done at a SID of 180 cms with field size of 900 cm2 (30 × 30). The patient thickness is 30 cm. Hence, the source of patient to distance is 180 – 30 = 150 cm

⎛ 90 , 000 ⎞ ⎛ 0.15 ⎞ ⎛ 900 ⎞ XS = ⎜ ⎟×⎜ ⎟×⎜ ⎟ 2 ⎝ 1.5 ⎠ ⎝ 100 ⎠ ⎝ 400 ⎠ = 135 mR/wk The exposure due to leakage radiation can not exceed 100 mR/hr at 1m from the source, for the maximum tube current and maximum kVp. This may be written as 100 mR/mAmax – hour at 1 m. This can also be expressed in mR/mA.min, by dividing by 60. Then the leakage radiation (XL) per week is the product of the leakage radiation per mA min and the work load (Assume maximum m A=5, W = 300 mA.min). 43

Textbook of Radiological Safety Maximum leakage = 100 mR/mA × 60 = 1.67 mR/mA min XL = (1.67 mR/mA min) × W (mA min/wk)

(7)

= (1.67/5 mA min) × 300 (mA min/wk) = 100.2 mR/wk The total weekly exposure (X) is due to primary, scatter and leakage radiations can be calculated, for an unshielded condition as follows: ⎛ exposure ⎞ ⎡⎛ X P X⎜ ⎟ = ⎢⎜ 2 ⎝ week ⎠ ⎢⎣⎜⎝ dpri

⎞ ⎛ X ⎟ U pri + ⎜ S 2 ⎟ ⎝ dsec ⎠

⎞ ⎛ XL ⎟+⎜ 2 ⎠ ⎝ dleak

⎞⎤ ⎟⎥ ⎠ ⎥⎦

(8)

For example as shown in the Fig. 3.5, if a wall is positioned at 2.5 m from the source, and it is 0.5 m from the chest stand, then dpri=2.5m, dleak=2.5m, dsec=0.5, and U =1/4, using this in the above equation,

Fig. 3.5: A model X-ray layout for barrier calculation

X =

⎡⎛ 90.000 ⎞ ⎛ 1 ⎞ ⎤ ⎛ 135 ⎞ ⎛ 100.2 ⎞ ⎢⎜ 2.5 2 ⎟ × ⎜ 4 ⎟ ⎥ + ⎜ 0.5 2 ⎟ + ⎜ 2.5 2 ⎟ ⎠ ⎝ ⎠⎦ ⎝ ⎠ ⎝ ⎠ ⎣⎝

= 4156 mR/wk If Xlimit is the regulatory limit of radiation exposure for an partial occupancy (corridor, T=1/4) area, then Xlimit = 2 (mR/wk) / T, for uncontrolled area, since T =1/4 =0.25 = 2 (mR/wk) / 0.25 = 8 mR/wk

44

Radiation Exposure Control The attenuation offered by the barrier is given by the relation Xlimit =X × e-μ ´ T

(9)

where the linear attenuation coefficient m = 0.693/HVL or μ = 2.303/TVL and T is the thickness of the attenuator. The HVL and TVL represents the half value layer and tenth value layer of the construction material of the barrier. The HVL and TVL are functions of the kVp and workload. The values of the same is given Table 3.6 for various diagnostic energies for lead and concrete Xlimit = X e-(2.303 × T/ TVL) Xlimit/X = e-(2.303 × T /TVL) by taking logarithm, and solving for T, the equation may be written as T = 0.434 × ln (X / Xlimit) × TVL By substituting the values of exposure (X) and exposure limits (Xlimit) from above and TVL of concrete for 125 kVp (Table 3.6) T = 0.434 × ln

4156 mR/wk × 66 = 179.1 mm 8 mR/wk

This may be approximated as 18 cm or 7 inch concrete thickness. If brick is used as construction material then 10.5 inches thickness will offer the same protection. Instead of concrete if lead is used, then TVL =0.93, for 125 kVp from the table (3.6), then T = 0.434 × ln (4156/8) × 0.93 = 2.52 mm. Usually lead is specified in weight per square foot (lb/sq.ft) and 1mm of lead thickness = 2

1 lb/sq ft. There is some practical difficulty in handling 2

lead greater than 2 lb/sqft, and hence that is the recommended minimal thickness. In this calculation, the attenuation offered by the screen-film is not accounted. Hence, the wall thickness is over estimated. Any way it is good to compensate increased workload of the future. Similar way the calculation is repeated for all the three walls (B, C, and D) and ceiling for the layout in Fig. 3.5. In multiple floor buildings, the floor also should be considered for barrier calculations. The calculated shielding should extend up to a height of 7 feet from the floor. All X-ray rooms need not be designed with concrete. For example, mammography unit operates within 35 kVp and 2 sheets of gypsum drywall (1.6 cm thickness) is sufficient, to achieve the required radiological safety.

45

Textbook of Radiological Safety Table 3.6: HVL &TVL Values of lead and concrete for various diagnostic X-ray energies Peak voltage, kV

HVL,Lead (mm)

TVL,Lead (mm)

50 70 100 125 150

0.06 0.17 0.27 0.28 0.30

0.17 0.52 0.88 0.93 0.99

HVL,Concrete TVL,concrete (mm) (mm) 4.3 8.4 16.0 20.0 22.4

15 28 53 66 74

Shielding Design for Computed Tomography Work Load In the case of CT scan all the walls in the room are secondary barriers, and the detector plays the role of primary barrier. The measured data from a individual CT scan is required to determine the amount of scattered radiation, arising from the gantry. Normally, these measured data are provided as exposure lines from the isocenter of the gantry on a per slice basis for a given mAs (Fig. 3.6). The work load of a CT scan is calculated from average number of patients per week, fraction of head verses body scans, and the average mAs per patient.

Fig. 3.6: CT scan layout: Exposure levels measured from the isocentre of the gantry in mR / hour

46

Radiation Exposure Control The workload of a head CT scan having 20 abdominal scan per day, 30 slices per scan with 250 mAs per slice can be calculated as follows. 50 % of the scans are pre and post contrast and 50 % are with out any contrast (assume 1.5 studies /patient) W = (20 patients /day) × (5 day /wk) × (30 slices /study) × (1.5 studies /patient) = 4500 slices /wk Shielding Calculation If each slice provides 0.08 mR exposure at a distance of 3.5 m from the iocentre, for 100 mAs, then the weekly exposure (X) is = 4500 slices × (0.08 /100mAs) × (250 mAs /slice) = 900 mR/week If the permissible limit is 2 mR /week, for a occupancy of T=1/4=0.25, then Xlimit = (2 mR /week) / 0.25 = 8 mR /week The required concrete shielding to reduce the exposure to 8 mR /week is T = 0.434 × ln (X / Xlimit) × TVL = 0.434 × ln (900 /8) × 66 = 135.2 mm or 13.5 cm or 5.31 inches If lead is used instead concrete, then T = 0.434 × ln (900 /8) × 0.93 = 1.90 mm This thickness is equal to 5 lb /sq ft. Similar calculation needs to done for other walls, floors and ceilings. Now a days spiral CT scanners are used with increased acquisition speed and efficiency, which demands high output X-ray tubes. These scanners provides additional radiation burden to the workers and public. For example, a 64 slice CT scan handles 150 patient /week, 5 rotation per patient with 250 mAs exposure per rotation. Assume the exposure level at the wall level is 0.13 mR for 100mAs. The corresponding weekly exposure is = 150 patient /week × (5 rotation /patient) × (0.13 mR /100 mAs) × (250 mAs / rotation) = 243.7 mR /week. FACILITY DESIGN FOR NUCLEAR MEDICINE The radiation exposure in nuclear medicine may vary from natural background to 100 R/hr. This is calculated by using the specific exposure rate constant (Γ) of the radionuclide, which is defined as the exposure rate in R/hr at 1 cm from 1mCi of the radionuclide. 47

Textbook of Radiological Safety Exposure rate (R/hr) = ΓA/d2

(10)

where A is the activity in mCi, and d is the distance in cm from a point source of radioactivity. Most of the low energy photons are attenuated by air and other intervening materials. Hence, the exposure rate constant do not account photon energy below certain levels. If the exposure rate constant accounts photon energy greater than 20 keV, then it is denoted as Γ20. Example 6: A vial contains 100 mCi of I-131, having a Γ20 =2.18 R-cm2/ mCi-hr. What will be the exposure at 100 cm, if a nursing staff sit for 1 hour. Exposure = (2.18 R-cm2 / mCi-hr) × (100 mCi) × (1 hr) × (1/100 cm2 ) = 0.0218 R = 21.8 mR In general, there is no need to shield the walls of the nuclear medicine laboratory. However tungsten, lead and lead glass are used as shielding materials to give personnel protection. These are used to design syringe shield, L-bench, lead pigs, transport container and waste dust bin etc. The Table 3.7, presents the exposure rate constants and HVL for various radioisotopes. Table.3.7: Exposure rate constants and HVL for various radionuclides Radionuclide

Γ20 (R-cm2 / mCi-hr)

Half value layer,Lead, mm

I-131 Mo-99 Tc-99m F-18 Ga-67 Cs-137

2.18 1.47 0.62 5.66 0.75 3.25

3.0 7.0 0.3 3.9 1.0 5.5

FACILITY DESIGN FOR RADIOTHERAPY 1. Workload: In radiotherapy, the workload is obtained by multiplying number of patient treatment per week with delivered dose for each patient. It is expressed in Gy/week. For example, if a linear accelerator treats 50 patients per day with dose of 4 Gy for each patient at 1m, then W = (4 Gy /patient) × (50 patient /day) × (5 days /week) = 1000 Gy /week

48

NCRP Report 49 suggests a workload figure of 1000 Gy/week based on a dose of 4 Gy at 1 m per patient, assuming a five day week for megavoltage facilities.

Radiation Exposure Control For dual energy machines the same workload figure of 1000 Gy/ week may be used. NCRP Report 51 suggests an assumed workload of 500 Gy/ week for the higher energy, with the remainder of the workload being attributed to lower energy X-rays or electrons. 2. Use Factor: A use factor (U) describes the different beam orientations used for treatment when calculating the required barrier thickness for each beam orientation. If conventional treatment techniques are to be used, NCRP Report 49 suggests a use factor of 1 for the floor with the beam pointing vertically down, and 0.25 for each wall and ceiling, if specific values are not available. These use factors may depend on the particular use of the facility and also on the energy used. For example, a facility performing a large number of total body irradiations may have a use factor greater than 0.25 for one wall, and lower for other walls. 3. Occupancy Factor: The occupancy factor (T) for an area should be considered as the fraction of time spent by a single person; who is there longest. It is most likely that the target group for shielding purposes will be non radiation workers employed by the hospital. The occupancy factor is best defined as the fraction of an 8 h day or 2000 h year for which a single individual may occupy a particular area. Occupancy factors are based on local regulations and the specific conditions at the facility under consideration. IDR and TADR The instantaneous dose rate (IDR), is the direct reading of the dosimeter that gives a reading in dose per hour, averaged over one minute. When calculating the required barrier shielding it is useful to calculate the expected IDR for comparison with direct measurement after the facility has been built and the treatment unit installed. The time averaged dose rate (TADR) is the barrier attenuated dose rate averaged over a specified time. TADR is proportional to IDR, and incorporates the W,U and dose output rate (DR0) of the unit. The TADR is estimated over 8 h (R8), by taking into account the workload and occupancy factor as unity. For megavoltage treatment units, although the unit may be in use for 8 h per day, it is likely that the total beam-on time per day will be much less. The TADR, or R8, is then determined from the IDR multiplied by the daily beam on time and then divided by the length of the working day: Daily beam on time TADR = IDR × (11) Length of working day or, It can also be expressed as follows: R 8 = IDR ×

Wd U 8 × DR o

where R8 is the TADR averaged over an 8 h day, in mSv·h–1 ;

(12)

49

Textbook of Radiological Safety IDR is in mSv·h–1 averaged over 1 min at a point 0.3 m beyond the barrier, with the machine operating at the dose output rate DR0 ; Wd is the daily workload defined at 1 m, in Gy for an 8 h day; U is the use factor (=1 for secondary barriers or the maze entrance); DR0 is the dose output rate at 1 m, in Gy·h–1 or Sv·h–1. From the above, the weekly TADR (RW) may be calculated, as follows: ⎛ WU ⎞ R W = IDR × ⎜ ⎟ ⎝ DR o ⎠

(13)

where W is the weekly workload, for 40 hours. The TADR 2000 is the time averaged dose rate estimated over 2000 h. This takes into account the workload, use factor and occupancy factors. According to guidance from the United Kingdom, if the IDR is less than 7.5 mSv·h–1 and the TADR is less than 0.5 mSv·h–1 or the TADR 2000 is less than 0.15mSv·h–1, the area need not be supervised. Primary Barriers Weekly Dose Rate Method Let P is the permissible dose equivalent limit for a given area in Sv/wk. If B is the required attenuation of the barrier to reduce the primary beam dose to P, then P=

WUT × B

( d + SAD )

2

or

B=

P ( d + SAD ) WUT

2

(14)

where d is the distance between the isocentre and the area of interest outside the primary barrier in meters, SAD is the source to axis distance in meters, W is the work load in Gy/week U is the use factor, and T is the occupancy factor. NCRP-51 has given the beam attenuation curves for various energies and barrier materials, under broad beam geometry. By referring these attenuation curves, the barrier thickness can be obtained. Fig. 3.7 gives the model attenuation curves for various electron beam energies with concrete of density 2.35 g/cc as barrier material. The thickness of concrete required can be determined from attenuation graphs, or by the use of TVLs. The number of TVLs required to produce this attenuation is determined from: Number of TVL (n) = log10 (1/B)

50

(15)

Radiation Exposure Control

Fig. 3.7: Attenuation curves for various electron beam energies under broad beam geometry, for concrete with density 2.35 g / cc as barrier material

The Table 3.8, presents the TVL values in meters for various construction materials, with variable beam energies. Then, the thickness of the barrier is calculated by using the equation S = T1+ (n – 1)Te

(16)

where, T1 is the first TVL and Te is the equilibrium or subsequent TVL of concrete as given in the NCRP report No.51. The Table 3.9 presents the first TVL (T1) and subsequent TVL(Te) for various construction material with respect to energy. Primary barrier

B=

P ( d + SAD )

2

WUT

Assume P = 0.12 mSv/week, d = 3 m, U = 0.25, T= 1, W = 384 × 103 mGy/wk B=

0.12 × ( 3 + 0.8 )

2

384 × 10 × 0.25 × 1 3

= 1.81 ×105

Number of TVL’s = log 10 (1/B) = log 10 (1/ 1.81 × 105) = 4.74 The TVL for 60Co in concrete (density 2350 kg·m–3) is 218 mm (Table 3.8). Therefore, the required thickness for the primary barriers is 4.74 × 218 =1033 mm.

51

Textbook of Radiological Safety Instantaneous Dose Rate Method The barrier thickness required to reduce the IDR to an acceptable level on the far side of the barrier is determined as follows: The attenuation required BIDR is given by: B IDR =

PIDR ( d + SAD )

2

(17)

DR 0

where PIDR is the instantaneous design dose limit, in Sv·h–1; d is the distance from the isocentre to the point of interest on the far side of the barrier, in metres; SAD is the source–axis distance (usually 1 m for linear accelerators); DR0 is the dose rate at the isocentre (1 m), in Gy·h–1. The number of TVLs of concrete is then determined from Eq.(15) in the same way as for the weekly dose rate method. Wall thicknesses determined for primary barriers will be more than adequate to shield against leakage and scattered radiation and no further calculations are required. Table 3.8: TVL values for various construction materials, with variable beam energies (Source: IAEA safety series 47) Co-60a 4MVb 6MVb 10MVb 15MVb 18MVb 20MVb 24MVb TVL for concrete (density 2350 kg . m-3) (in mm) Primary beam 218 290 343 389 gamma/X-rays

432

445

457

470

305

330

330

343

356

TVL for steal (density 7800 kg . m-3) (in mm) Primary beam 71 91 98 105 gamma/X-rays

108

111

111

107

85

87

87

88

89

TVL for lead (density 11360 kg . m-3) (in mm) Primary beam 41 53 55 56 gamma/X-rays

57

56

55

52

47

47

49

51

Leakage gamma and X-rays (900)

Secondary beam gamma/X-rays

Secondary beam gamma/X-rays

218

69

40

254

79

47

279

80

45

46

a: NCRP report 49(1976), b: Varian Associates, Nelson and LaRiviere (1984).

Width of the Primary Wall

52

The primary barrier width is made equal to the maximum field size at the barrier plus 1 foot (0.305 m) on either side to prevent radiation from leaking through the secondary barrier that abuts the primary. Most of the linear accelerator has 40 cm × 40 cm as maximum field size at 1m from the target.

Radiation Exposure Control Table 3.9: Tenth value layers for primary barriers, for concrete(2.35 g/cm3), steel (7.87 g/ cm3) and lead (11.35 g/cm3) Vs beam energies (NCRP 151,2005) End point energy (MV) 6

10

15

18

20

25

Co-60

Shield material

T1 (cm)

Te(cm)

Concrete Steel Lead Concrete Steel Lead Concrete Steel Lead Concrete Steel Lead Concrete Steel Lead Concrete Steel Lead Concrete Steel Lead

37 10 5.7 41 11 5.7 44 11 5.7 44 11 5.7 46 11 5.7 49 11 5.7 21 7 4

33 10 5.7 37 11 5.7 41 11 5.7 41 11 5.7 44 11 5.7 46 11 5.7 21 7 4

When the collimator is rotated to 45 °, the above dimension become equal to its diagonal (56.6 cm). Then the horizontal barrier width (W), required is given by W = 0.566 d + (0.305 × 2) = 0.566 d + 0.61

(18)

where d is the distance from the source to the barrier. Secondary Barrier for Scattered Radiation If Bp is the transmission factor required to reduce the scattered dose to an acceptable limit P in the area of interest, then BP =

P × dsca 2 × dsec 2 αWT × ( F / 400 )

(19)

where P, W and T have the same meaning as above, dsca is the distance from the radiation source to the patient, in m. dsec is the distance from the patient to the point of interest, in m. α is the scatter fraction defined at dsca. The scatter primary ratio (α) is dependent on the energy of the X-ray beam and the scattering angle. 53 F is the field area incident on the patient, in cm2.

Textbook of Radiological Safety The α is the reflection coefficient of the barrier material, which is defined as the fractional scatter at 1m from the scatter, for a beam area of 400 cm2 incident at the scatter. It is the ratio of the scattered dose to the incident dose, which varies with scattering angle and beam quality. Table 3.10 gives the α values for different scattering angles. Radiation scattered by a patient or phantom is usually less than 0.1% of the incident radiation per 0.1 m2 area irradiated. For large scatter angles, the energy of the scattered radiation will be degraded and the protection designed against leakage radiation should provide adequate protection against scattered radiation from the patient. However, when the scatter angle is small, patient scatter should not be ignored. For small scatter angles (10°), the value of the scatter fraction a may be as high as 0.0178 for 24 MV X-rays and the scattered photons have energies close to that of the incident beam. The use factor for the secondary barrier is unity. The required thickness is obtained from attenuation curves of the NCRB-51 report. Table 3.10: The values of α for different scattering angles for Co-60 and 6MV beams (IAEA safety series 47) Scattering angle 10 20 30 45 60 90 135 150

Co-60 (10-3)

6MV(10-3)

10MV(10-3)

11 80 6.0 3.7 2.2 0.91 0.54 0.15

16.8 11.5 5.36 2.97 1.74 0.727 0.488 0.328

16.9 10.3 6.73 3.25 1.84 0.714 0.370 0.316

18MV(10-3) 24MV(10-3) 24.3 11.7 7.13 3.05 1.42 0.375 0.259 0.226

27.4 12.7 7.21 3.06 1.37 0.353 0.233 0.212

Secondary Barrier for Scattered Radiation, When the Primary Beam Strikes the Wall The barrier transmission factor (BW) needed to shield against scattered radiation when the primary beam strikes a wall is given by the following equation: BW =

Pdw 2 dr 2 αAWUT

(20)

where dw is the distance from the radiation source to the scattering surface (wall), in m; dr is the distance from the scattering surface (wall) to the point of interest, in m; α is the wall reflection coefficient, which depends on the wall material, 54 scattering angle, and beam energy,

Radiation Exposure Control A is the field area projected on the scattering surface (wall), in m2. The photons scattered by the wall and by the patient are of about the same energy. If the thickness required to shield from patient scatter is different from that needed to shield from wall scatter by one TVL or more, use the larger thickness, otherwise, use the larger thickness and add onehalf value layer (HVL). If the thickness required to protect from leakage differs from that required to protect from scatter by less than one TVL, use the greater thickness and add one HVL of shielding material for the energy of the leakage radiation. If the two thicknesses for leakage and scatter protection differ by more than one TVL use the greater thickness. Ceiling The roof section that can be struck directly by the radiation beam must be a primary barrier and the formulae used to determine the required thickness are the same as above. The design dose limit for the roof will depend on the location of the bunker. If it is a single storey building, then the only consideration may be the limitation of access to the roof space. However, if the building is overlooked, then the effect of skyshine must be considered which may result in the irradiation of nearby buildings. If there is a nuclear medicine department nearby, then it should be noted that gamma cameras and possibly other imaging equipment are particularly sensitive to low levels of radiation that can affect certain patient investigations. If the building has further floors above the bunker, then consideration should be given to locating a storage room or plant room immediately above the bunker (A plant room is used to house the chiller unit for the linear accelerator or heating and ventilation system plant). A storage room or plant room will have limited occupancy and access can be restricted, thus allowing a greater design dose limit than if an office was placed directly above the bunker. Secondary Barrier for Leakage Radiation For a linear accelerator, national and international protocols state that the leakage from the treatment head must not exceed 0.5% of the primary beam, outside the useful beam at 1 m from the path of the electrons between the gun and target window and averaged over 100 cm2. In the plane of the patient, the leakage must not exceed an average of 0.1% and a maximum of 0.2% over a 2 m radius measured from the beam central axis. In general, manufacturers have protected their machines to better than 0.1%, and it would be reasonable to assume this value when determining the required secondary barrier thickness. If BL is the transmission factor required to reduce the leakage dose to the permissible limit P, then BL =

1000Pds 2 WT

(21) 55

Textbook of Radiological Safety where dS is the distance from the isocentre to the area of interest for secondary barrier. The leakage radiation is present, whenever the machine is switched on and hence the use factor is unity (U=1). Since the use factor is unity the average position of the treatment head is taken to be at the isocentre so the distance ds is measured from the isocentre. In the case of mega voltage machine > 500 kVp, the leakage is 0.1% through the source housing, which is equal to 1/1000. The quality of the leakage radiation is the same as primary. Hence, the transmission curves of the primary can be used for leakage radiation. Concrete, lead or steel are used as barrier materials, which depends on structural and spatial considerations. Since concrete is relatively cheep, it is commonly used as a barrier material to design walls and ceilings. High density concrete (3.4-3.5 g/cc), lead and steel are recommended as barriers, whenever there is a scarcity of space. A number of different materials such as magnetite, barites, iron scrap and hematite may be mixed with concrete. The physical properties of common shielding materials are given in Table 3.11. The leakage radiation barrier thickness is always greater than scattered radiation, since leakage radiation is more penetrating in nature. In the case of low energy, the difference in barrier thickness is very small. An optimally designed primary barrier will ensure adequate protection against scattered and leakage radiation. Table 3.11: Physical properties of common shielding materials Shielding material Concrete High density concrete Low-carbon steel Lead

Density, g/cm3

Atomic number

2.35 3.4-3.5 7.90 11.4

11 26 26 82

Worked Examples Example 7: A 60Co treatment facility, has 40 patients/day (8h) and the dose delivered per patient at the isocentre is 3 Gy. The facility is used for 5 days per week. The source specification is 0.8 Gy/min at 1 m, and the isocentric distance of the treatment unit (SAD) is 80 cm. Calculate the DR0, workload, and beam on time per day The dose rate at the isocentre = 0.8 × (100/80)2 × 60 =75 Gy/h. The dose rate at 1 m (DR0) = 0.8 × 60 = 48 Gy/h Workload = 40 × 3 × 5 = 600 Gy/week at the isocentre (at 80 cm) or

56

= 600 ×

0.8 2 = 384 Gy/week at 1 m. 1.0 2

Radiation Exposure Control The total dose delivered at the isocentre per day = 40 × 3 =120 Gy. Total beam-on time per day =120 × 75 = 1.6 h. Example 8: Calculate the primary barrier thickness by using weekly workload method and IDR (Assume effective dose limit as 20 mSv per year). P

= 0.40 mSv /week,d= 3 m,

W

= 384 Gy /week at 1 m = 384 ×103 mGy/week

SAD = 0.8 m, U=0.25, T=1 DR0 = 48 Gy /h = 48 × 106 μGy /h B=

P ( d + SAD )

0.4 ( 3 + 0.8 )

2

=

WUT

2

384 × 10 × 0.25 × 1 3

= 6.01 × 10 −5

Number of TVLs = log 10 [1/ (6.01 × 10-5)] = 4.22 The TVL for 60Co in concrete (density 2350 kg·m–3) is 218 mm and hence the required thickness for the primary barriers is (4.22 × 218) 920.2 mm. For this barrier thickness the IDR beyond the barrier is determined. IDR =

Β × DR 0

( d + SAD )

2

=

6.01 × 10 −5 × 48 × 10 6

( 3 + 0.8 )

2

= 199.7 μ Sv/h

Example 9: Calculate the primary barrier thickness by using weekly workload method (equation 14). (Assume P = 0.3 mSv/year as limit for design purposes). P

= 0.3 mSv/50 week= 6 μSv/week

d

= 3 m, W= 384 Gy/week at 1 m

SAD = 0.8 m U B=

= 0.25, T=1 P ( d + SAD ) WUT

2

=

6 × ( 3 + 0.8 )

2

384 × 10 6 × 0.25 × 1

= 9.03 × 10 −7

Number of TVLs = log 10 [1/ (9.03 × 10-7)] = 6.04 Therefore, a (6.04 ×218)1317 mm thick concrete primary barrier is required to shield a public area with an occupancy of 1. For a public area with occupancy factor of 0.5, a similar calculation gives the barrier thickness of 1252 mm. For a public area with an occupancy factor of 0.2, the barrier requirement is 1165 mm.

57

Textbook of Radiological Safety Example 10: For leakage radiation from the treatment head, the manufacturer’s specification should be used. There may be two values of leakage radiation quoted by the manufacturer, one when the source is in the safe position and one when the source is exposed for treatment; the larger value should be used in the shielding calculations. This value is usually less than the 0.1% (1/1000) of the primary radiation that is allowed. To determine the required barrier thickness, Eq.(21)is used. In this example: ds the distance from the isocentre to just outside the secondary barrier = 2.6 m P the design limit for a public area is 20 μSv/week (1mSv/50 week) T, the occupancy is 1. B=

1000 × 20 × 2.6 2 = 3.52 × 10 −4 384 × 10 6 × 1

Number of TVLs = log 10 [1/ (3.52 × 10–4 )] = 3.45 The thickness of concrete required is 3.45 × 218 = 752.1 mm IDR =

DR 0 × B 48 × 10 6 × 3.52 × 10 −4 = = 2.49 μ Sv/h 2 1000 × dS 1000 × 2.6 2

Example 11: The barrier thickness necessary to shield against radiation scattered by the patient is determined from Eq. (19); P is 20 μSv/week dsca the isocentric distance is 0.8 m; dsec has the same value as dS in the previous calculation, 2.6 m; α the scatter fraction for 90 degree scatter is 0.0009 per 400 cm2 of area irradiated; F is the maximum field area incident on the patient (20 cm × 20 cm) =400 cm2

58

BP =

P × dsca 2 × dsec 2 αWT × ( F / 400 )

BP =

20 × 0.8 2 × 2.6 2 = 2.5 × 10 −4 0.0009 × 384 × 10 6 × 1 × ( 400 / 400 )

Number of TVLs = log 10 [1/ (2.5 × 10–4 )] = 3.60

Radiation Exposure Control This is similar to the number of TVLs required to shield against leakage radiation. For 90 degree scatter the energy of the scattered radiation will be degraded and the protection designed for the leakage radiation should provide adequate protection against radiation scattered from the patient. Example 12: It is proposed to design a 6 MV linear accelerator facility for a work load of 1000 Gy /week at the isocentre. The location of a particular primary walls is at 3.6 m from the isocentre. Assume the use factor and occupancy factor are unity. The permissible limit (P) is 1 mSV per year= 1mSv / 50 week = 0.02 mSv/ week SAD = 1 m, d = 3.6 m W = 1000 Gy/week U = 1, T = 1 B=

0.02 × 10 −3 ( 3.6 + 1)

2

1000 × 1 × 1

= 0.423 × 10 −6

Number of TVLs = log 10 [1/ (0.423 × 10–6 )] = 6.37 The actual thickness required (S) = T1 + (n – 1)Te = 0.37 + (6.37 – 1) × 0.33 = 2.14 m Since, the first TVL(T1) and subsequent TVL (Te) for 6MV is 0.37 and 0.33 m, from Table 3.9. Example 13: Find the width of a primary barrier, which is 3.6 m away from a 6 MV Linear accelerator. By using equation 18, W = 0.566 d + 0.61 = (0.566 × 3.6) + 0.61 = 2.03 + 0.61 = 2.64 m Facility Design for Brachytherapy To determine the required attenuation of the primary barriers, Eq. (14) is used. For brachytherapy the workload W is based on the dose delivered per treatment and the number of treatments: W = RAKR × A × t × n

(22)

59

Textbook of Radiological Safety where RAKR is the reference air kerma rate for a source of unit activity; A is the total activity (activity per source × number of sources); t is the average duration of treatment in hours; n is the number of treatments per week. Using the AAPM Report 21 specifications, the workload may be represented by: W = Sk × t × n where Sk is the air kerma strength of the source in units of U or µGym2·h–1. Similarly, the dose rate D0 will be given by: D0 = RAKR × A

(23)

or, using the AAPM Report 21 specifications: D0 = Sk For brachytherapy the sources are not collimated so the use factor U will always be unity. A modified version of Eq. (14) for brachytherapy shielding may be written as: B=

P × d2 RADR × A × t × n × T

or

B=

P × d2 Sk × t × n × T

(24)

where P is the design limit; d is the distance, in m, from the exposed source position to the point of interest outside the barrier; T is the occupancy of the area outside the barrier. The RAKR values of different Brachytherapy sources are given in Table 3.12. Table 3.12: Physical data of Brachytherapy radionuclides Nuclide

Mean photon energy(MeV)

Half-life

RAKR (µGy MBq-1 . m2 . h-1)

Co-60 I-125 Cs-137 Ir-192 Au-198 Ra-226

1.25 0.028 0.662 0.37 0.42 0.78

5.27 y 60.1 d 30.0 y 74.0 d 64.7 h 1600 y

0.308 0.034 0.077 0.111 0.056 0.195

Unlike megavoltage bunkers, brachytherapy rooms are not used so regularly. Their use is often limited by the number of operating room sessions available for placing the source applicators in the patient. Consequently, basing the shielding design on an annual dose limit may result in high IDRs outside the barriers. This may necessitate these areas 60 being designated as controlled areas during the course of the treatment if

Radiation Exposure Control the IDR exceeds 7.5 mSv·h–1. It is therefore recommended that the IDR be assessed (based on the maximum number of sources normally used) and also the maximum dose rate (based on the maximum number of sources available) before finalizing the shielding design. The Table 3.13, presents the minimum concrete thickness required to reduce the dose rate to 7.5 and 2.5 mSv/h at a distance of 3 m from the source. Table 3.13: Typical concrete barrier thickness required at 3 m from the radiation source Type

Radionuclide

Activity, GBq

MDR afterloading HDR after loading HDR after loading

Caesium-137 Iridium-192 Cobalt-60

22.2 370 185

Thickness (mm) to reduce the dose rate to 7.5 µSv/h 2.5 µSv/h 280 440 680

360 510 770

Example 14: The Caesium manual after loader has 5 sources of each 100 mCi (3.7 GBq), to be used for gynaecological treatments. The reference air kerma rate (RAKR) for 137Cs is 0.077 μGy·MBq–1·m2·h–1.. The intended workload is 5 treatments per week. The shielding design will be based on the use of 5 sources per patient with a total activity of 18.5 GBq (0.5 Ci). The average treatment duration is 30 h to deliver an absorbed dose of 30 Gy to the prescription point. The weekly workload is obtained from Eq.(22): W = 0.077 × 18.5 × 103 × 30 × 5 = 2.13 × 105 μGy.m2 The design limit is 20 mSv /week for a public area (T=0.1) at 3.5 m from the treatment position of the sources. The required transmission through the barrier is determined from Eq. (24): B=

20 × 3.5 2 = 1.15 × 10 −2 2.13 × 10 5 × 0.1

The number of TVLs required is log 10[1/(1.15) × (10-2)] = 1.93. The TVL for caesium 137 for concrete is 175 mm and the total thickness of concrete required is 1.93 × 175 mm =337.7 mm. Example 15: A HDR Brachytherapy machine has 10 Ci (370 GBq) activity of Iridium -192 source, the average photon energy is 0.38 MeV and the RAKR is 0.111 mGy·MBq–1·m2·h–1.. The intended workload is 20 treatments per week. The shielding design will be based on the use of 10 Ci sources per patient with average treatment duration of 5 min (0.08 h), to deliver an absorbed dose of 8 Gy to the prescription point. The weekly workload is obtained from Eq. (22): W = 0.111 × 370 × 103 × 0.08 × 20 = 6.57 × 104 μGy.m2

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Textbook of Radiological Safety The design limit is 20 μSv/week for a public area (T=0.1) at 3.5 m from the treatment position of the sources. The required transmission through the barrier is determined from Eq. (24): B=

20 × 3.5 2 = 3.7 × 10 −2 6.57 × 10 4 × 0.1

The number of TVLs required is log 10[1/(3.7 × 10-2)] = 1.43 The TVL for Iridium -192 for concrete is 152 mm and the total thickness of concrete required is 1.43 × 152 mm =217.3 mm. Example 16: The HDR unit contains 20 60Co sources each of 18.5 GBq (500 mCi). The reference air kerma rate (RAKR) for 60Co is 0.308 mGy·MBq– 1 ·m2·h–1. The intended workload is 30 treatments per week. Calculate the dose rate and the barrier thickness in concrete. The dose rate is calculated by using the equation 23; Do = 0.308 × 18.5 × 103 × 20 = 113,960 mGy·m2·h–1 To reduce the dose rate to the design limit is 7.5 mSvh–1 at 3.5 m (UK data) B=

7.5 ( 3.5 )

2

113 , 960

= 8.1 × 10 −4

The number of TVLs required is log 10[1/(8.1 × 10-4)] = 3.09 The TVL for Cobalt -60 for concrete is 218 mm and the total thickness of concrete required is 3.09 × 218 mm =673.6 mm. BIBLIOGRAPHY 1. AERB safety code: Brachytherapy sources equipment and installations. SC/ MED-3. 1988. 2. AERB safety code: Medical diagnostic X-ray equipment and installations. SC/MED-2(Rev.1)2001. 3. AERB safety code:Telegamma therapy equipment and installations. SC/MED1.1986. 4. FM Khan. The Physics of Radiation therapy, (3rd edn.) Lippincott Williams & Wilkins 2003. 5. Jerrold TB, Seiber JA, Edwin ML, John MB. The essential physics of medical imaging, (2nd edn.), Lippincott Williams & Wilkins 2002. 6. NCRP. Medical X-ray, electron beam and gamma ray protection for energies up to 50 MeV. Report No. 102. Bethesda, MD: National Council on Radiation Protection and Measurements, 1989. 7. NCRP. Radiation protection design guidelines for 0.1-100 MeV particle accelerators facilities. Report No.51. Washington DC: National Council on Radiation Protection and Measurements,1977.

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Radiation Exposure Control 8. NCRP report 151.Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities. National Council on Radiation Protection and Measurements 7910 Woodmont Avenue, Suite 400/Bethesda, MD 20814-3095. 9. NCRP. Structural shielding design and evaluation for medical use of X-rays and gamma rays of energies up to 10MeV. Report no.49. Washington,DC: National Council on Radiation Protection and Measurements, 1976. 10. Patton HM. Shielding Techniques for radiation oncology facilities, (2nd edn.) Medical physics publishing. www.medical physics.org 2002. 11. Radiation protection in the design of radiotherapy facilities, Safety reports series No. 47. International Atomic Energy agency Vienna, 2006.

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Chapter

4

Planning of Radiological Facility

GENERAL GUIDELINES While planning a radiation facility, consideration should be given to radiation safety, economy and convenience. The following features are very important while planning a radiation installation either it is diagnostic radiology or Nuclear medicine or Radiotherapy facility. 1. Location: The site or room should be located as far away as feasible from areas of high occupancy and general traffic, such as maternity and pediatric wards and other departments of the hospital that are not directly related to radiation and its use. It should be preferably at the extreme end of the hospital and be easily accessible to various departments of the hospital. 2. Layout: The layout of rooms should aim at providing integrated facilities so that handling of radiation equipments and related operations can be conveniently performed with adequate protection. The installation should permit safe and easy transport of equipments and nonambulatory patients. 3. Room size: The room must be spacious enough to permit the radiation equipment and accessories, use and servicing of the equipment with safety and convenience. It should facilitate the wheeling in of patients in and around the couch of the unit. Proper grouping of the rooms comprising the installation should be done bearing in mind their dependence on each other. 4. Shielding: Appropriate structural shielding shall be provided for the walls, ceiling, floor, doors and windows, so that the doses received by the occupational workers and members of the public are kept to a minimum and shall not exceed the annual effective doses as prescribed by the competent authority. The current limits are 30 mSv and 1 mSv for the workers and the members of the public. 5. Doors: The number of doors for entry and windows should be kept minimum. It should permit safe and easy transport of equipment and nonambulatory patients. The doors shall provide the same shielding as that of the adjacent walls, in case persons are likely to be present in front of them, when the machine is energized. 6. Openings and ventilation: Unshielded openings, if provided in the room for ventilation or natural light must be located above a height of 2 m from the ground or finished floor level outside the room.

Planning of Radiological Facility 7. Equipment layout: The radiation equipment must be installed in such a way that in normal use the useful beam is not directed towards control panel, doors, windows, dark room and areas of high occupancy. The useful beam should preferably be directed towards unoccupied areas. Sufficient area should be left all around the couch for safe and free movements of equipment/ trolley, staff and service personnel. 8. Interlock: Suitable electrical interlocks between door, equipment and control panel must be provided, wherever it is necessary. 9. Warning light and placard: A suitable warning signal such as the red light must be provided at a conspicuous place outside the room and kept ON when the unit is in use, to prevent entry of persons not connected with the examination or treatment. An appropriate warning placard must also be posted outside the room entrance or door. 10. Air conditioning: The treatment room may be airconditioned to control temperature, pressure and humidity. It will ensure long-term, trouble free, safe operation of the equipment. Spilt AC with sufficient capacity to suit the room size is preferable than window AC’s. 11. Waiting area: In order to avoid the crowding of patients and relatives near the entrance door, waiting area must be provided outside and adjacent to the equipment room. It should have sufficient area to match the patient workload with toilet facility. 12. Emergency and trolly bay: There should be exclusive room for handling emergency. In addition, space must be provided to accommodate trolly and wheel chairs etc. ESTABLISHING A DIAGNOSTIC X-RAY FACILITY 1. Room size: The room housing an X-ray unit shall be not less than 18 m2 for general purpose radiography and conventional fluoroscopy equipment. The size of the room housing the gantry of the CT unit shall not be less than 25 m2. Also, not more than one unit of any type shall be installed in the same room, and no single dimension of these X-ray rooms shall be less than 4 m. In the case of mammography, the room size shall be not less than 10 m2, and no single dimension of the room shall be less than 3 m. A typical layout for a diagnostic X-ray department is shown in Fig. 4.1. 2. Wall thickness: If the X-ray installation is located in a residential complex, it shall be ensured that i. Walls of the X-ray rooms on which primary X-ray beam falls is (are) not less than 35 cm or 14 inch thick brick or equivalent, ii. Walls(s) of the X-ray room on which scattered X-ray fall is (are) not less than 23 cm or 9 inch thick brick or equivalent, and iii.There is a shielding equivalent to at least 23 cm or 9 inch thick brick or 1.7 mm lead in front of the door(s) and windows of the 65 X-ray room to protect the adjacent areas, either by general public

Textbook of Radiological Safety

3.

4.

5. 6. 7.

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or not under possession of the owner of the X-ray room. The density of the normal masonary brick is considered as 1.6 g/cc. iv.The ceiling must have a thickness of concrete (density 2.35 g/cc), not less than 6 inch or 13.5 cm. Control room: For equipment operating at 125 kV or above, the control panel must be installed in a separate control room located outside but contiguous to the machine room and provided with appropriate shielding, direct viewing (1.5 mm lead equivalence) and oral communication facilities between the operator and the patient. The X-ray units operating below 125 kVp in diagnostic radiology are exempted from the above class and may be located away from the primary beam, inside a stationery /mobile protective barrier. The protective barrier should have sufficient lead equivalence (1.5 mm). Both control console and machine can be housed in the same room. Doors: Doors to be lined with 1.5 mm thick lead sheet with proper overlapping at the joint and junction and wall of 9 inch thickness of brick and ceiling of 6 inch of concrete. Viewing window: Lead glass of suitable dimensions are provided as viewing windows of 1.5 mm thick lead equivalents. Mobile protective barrier: Control panel should be kept behind the mobile protective barrier (MBP) of thickness 1.5 mm lead equivalence. Dark room: The dark room should be located in such a way that the primary beam is not directed on it. Appropriate shielding must be provided for the dark room to ensure that undeveloped X-ray films

Fig. 4.1: A Typical layout of a diagnostic X-ray department

Planning of Radiological Facility stored in it will not be exposed to more than an air kerma rate of 10 μGy per week. 8. Placard: A warning placard as shown in the Fig. 4.2, must be posted outside the room entrance or door. 9. Approval: Two copies of the X-ray room layout drawn to scale 1:50, are to be sent for approval to Head, Radiological Safety Division, Atomic Energy Regulatory Board, Niyamak Bhavan, Anusaktinagar, Mumbai-400094 from radiation safety point and along with a required fee in the form of demand draft for an appropriate amount, drawn in favour of pay and accounts officer, AERB, payable in Mumbai, towards charges for approval of the layout. Mobile units does not require plan approval unless they are used as fixed units. Controlled and Uncontrolled Areas A controlled area is a limited access area in which the occupational exposure of personnel to radiation is under the supervision of an individual in charge of radiation protection. This implies that access, occupancy and working conditions are controlled for radiation protection purposes. In facilities that use X-rays for medical imaging, these areas are usually in the immediate areas where X-ray equipment is used, such as X-ray procedure rooms and X-ray control booths or other areas that require control of access, occupancy and working conditions for radiation protection purposes. The workers in these areas are primarily radiologists and radiographers who are specifically trained in the use of ionizing radiation and whose radiation exposure is usually individually monitored. Uncontrolled areas for radiation protection purposes are all other areas in the hospital or clinic and the surrounding environment. Uncontrolled areas are those occupied by individuals such as patients, visitors to the facility, and employees who do not work routinely with or around radiation sources. Areas adjacent to but not part of the X-ray facility are also uncontrolled areas.

Fig. 4.2: X-radiation warning sign

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Textbook of Radiological Safety GENERAL RADIOGRAPHY INSTALLATION These X-ray units are operated in the range of 50-150 kVp applied voltage. Objects that are irradiated are considered as primary barriers. Additional shielding must be provided for the wall behind the chest stand. Provisions are made to observe and communicate with the patient on the table. The operator should be in the control area. The shield at the control (protective barrier) must be a permanent /mobile one with 2.1 m height. It should not be used as primary barrier. The viewing window at the control barrier must be 45 × 45 cm size and centered. A typical model plan is shown in the Fig. 4.3 Area: 18 sq.m. (Not to scale) MPB: Mobile Protective Barrier of 1.5 mm lead equivalence V W: Viewing window of 1.5 mm lead equivalence W: Window or ventilator at a height of 2.0 m from the finished floor level outside the X-ray room. D: Entrance door lead lined with 1.0 mm of lead sheet with proper overlapping at all the joints. Walls of minimum 23 cm thick brick and ceiling 15 cm concrete.

Fig. 4.3: Model plan for a general radiography room

FLUOROSCOPY INSTALLATION Fluoroscopic imaging systems are usually operated at potentials ranging from 60 to 120 kVp. A primary barrier is incorporated into the fluoroscopic 68 image receptor. Therefore, a protective design for a room containing only

Planning of Radiological Facility a fluoroscopic unit, consider only secondary protective barriers against leakage and scattered radiations. However, provisions are made with primary barriers so that the function of the room can be changed at a later date without the need to add additional shielding. Most modern fluoroscopic X-ray imaging systems also include a radiographic X-ray tube. The shielding requirements for such a room are based on the combined workload of both units. A typical model plan is shown in the Fig. 4.4. Area: 7.5 × 5 =37.5 sq.m. (not to scale) V W: Viewing window of 1.5 mm lead equivalence W: Window or ventilator at a height of 2.0 m from the finished floor level outside the x-ray room D: Entrance door lead lined with 1.5mm of lead sheet with proper overlapping at all the joints and junctions. Walls of minimum 23 cm thick brick and ceiling 15 cm concrete. MAMMOGRAPHY INSTALLATION Mammography units are typically operated between 25-30 kVp. The walls are constructed with bricks or gypsum wall board. Adequate protective barrier of lead acrylic or lead glass are incorporated into dedicated mammography units. Doors need special attention as they offer poor attenuation than Brick or gypsum wall board. Gypsum wall board may contain voids and non uniform areas. Hence, higher thickness of gypsum wall board is recommended than that calculated. A typical model plan is shown in the Fig. 4.5.

Fig. 4.4: A typical layout of a fluoroscopy installation

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Textbook of Radiological Safety

Fig. 4.5: Model plan for a mammography room

Area: 3 × 5.5 =16.5 sq.m (Not to scale) W: Window or ventilator at a height of 2.0 m from the finished floor level outside the X-ray room. D: Entrance door lead lined with 1.0 mm of lead sheet with proper overlapping at all the joints. Walls of minimum 23 cm thick brick and ceiling 15 cm concrete. COMPUTED TOMOGRAPHY INSTALLATION Computed tomography (CT) employs a collimated X-ray fan-beam that is intercepted by the patient and by the detector array. Consequently, only secondary radiation is incident on protective barriers. The operating potential, typically in the range of 80 to 140 kVp, as well as the workload are much higher than for general radiography or fluoroscopy. Due to the potential for a large amount of secondary radiation, floors, walls and ceilings need special consideration. Additionally, scattered and leakage radiations from CT systems are not isotropic. The radiation levels in the direction of the gantry are much less than the radiation levels along the axis of the patient table. A typical model plan is shown in the Fig. 4.6. Area: 7 × 7.5 =52.5 sq.m. (not to scale) V W: Viewing window of 1.5 mm lead equivalence 70

Planning of Radiological Facility

Fig. 4.6: Model plan for a computed tomography room

W: Window or ventilator at a height of 2.0 m from the finished floor level outside the X-ray room D: Entrance door lead lined with 1.5mm of lead sheet with proper overlapping at all the joints and junctions. Walls of minimum 23 cm thick brick and ceiling 15 cm concrete. ESTABLISHING A NUCLEAR MEDICINE FACILITY General Guidelines 1. Location: The installation should be located in a relatively unfrequented part of the building so that access to the area can be easily controlled. It shall be located away from high patient or public occupancy areas and sources of intense radiation. Fire hazard potential should be minimum in the area chosen. The location of the installation or the facilities provided be such that the possibilities for spread of both surface and air-borne contamination are minimal. The location should be chosen that the minimum expenditure on shielding, radiation levels can be effectively maintained with permissible limits in the immediate vicinity. 2. Hot labs and radioactive storage areas should be located away from other busy work areas, public corridors, secretarial offices and away from imaging and low level counting rooms. 3. Areas of high activity and contamination shall be demarcated by physical barriers. Active areas shall be arranged in increasing order of 71 the activity with entrance from lowest active area.

Textbook of Radiological Safety 4. Walls, floor and doors of the active areas shall have hard, washable, nonporous and leak proof covering. Work surfaces shall be covered with nonporous and non reactive material. 5. Work benches should be sufficiently sturdy to support lead shielding. 6. Wash basins and sinks should be conveniently available where unsealed radioactive materials are handled. It is desirable that the sinks in hot labs have foot or elbow operated taps. 7. Plumbing shall provide direct flow of liquid effluents from active areas either directly to the delay tank or to the ultimate discharge point. Drain pipes and delay tank shall be leak proof and corrosion –resistant. 8. The laboratory design should permit separate storage of glassware and work tools (tongs, stirring devices) not used with radioactive materials to prevent needless contamination or mixture with similar items used with radioactive preparations. 9. Ventilation system shall be of once-through type with unidirectional air flow from areas of lower activity to higher activity. The exhaust from fume hoods shall be let out directly into the open after filtering. 10. Air conditioning is essential to maintain a clean, dust free and dry environment for electronic instruments that are sensitive to heat and moisture changes; high humidity is bad for electronic components, causing corrosion as well as current leakage. Instruments must be housed in an air conditioned environment, and a dehumidifier may be needed to maintain humidity at about 50%. 11. Running hot and cold water must be available. 12. Warning light and placard: A suitable warning signal such as the red light must be provided at a conspicuous place outside the room and kept ON when the unit is in use, to prevent entry of persons not connected with the examination or treatment. An appropriate warning placard must also be posted outside the room entrance or door (Fig. 4.7A) . Storage containers shall be posted with a different placard (Fig. 4.7B).

Fig. 4.7: Placard for storage radiation and containers rooms

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13. Planning and approval of nuclear medicine laboratory: Two copies of the layout of the nuclear medicine laboratory (drawn to scale 1:50)

Planning of Radiological Facility indicating the various rooms along with their dimensions, positions of doors, windows, exhausts, along with fume hoods, workbenches, sinks and other details should be sent to the Head, Radiological safety division, AERB, Niyamak Bhavan, Anus akti Nagar, Mumbai-400 094. A site plan (drawn to scale 1:500) indicating the location of the nuclear medicine laboratory and the occupancies around it including those above the ceiling and below the floor, if any, should also be sent to AERB. Categorization In the past, consideration was given to the categories of nuclear medicine ranging from simple imaging or in vitro laboratories to more complex departments, performing a full range of in vitro and in vivo procedures. These departments also involved in advanced clinical services, training programs, research and development. Now a days all assays (radioassays or enzyme linked immunosorbent assays (ELISAs) are done in biochemistry laboratories, whereas nuclear medicine departments are involved largely in diagnostic procedures, radionuclide therapy and nonimaging in vitro tests including RIA’s. The level of nuclear medicine services is categorized according to three levels of need: Level 1 This level is appropriate where only one gamma camera is needed for imaging purposes. The radiopharmaceutical supply, physics and radiation protection services are contracted outside the centre. Other services, such as radiology, cover receptionist and secretarial needs. A single imaging room connected to a shared reporting room should be sufficient, with a staff of one nuclear medicine physician and one technologist, with backup. This level is appropriate for a private practice. Level 2 This level is appropriate for a general hospital where there are multiple imaging rooms in which in vitro and other nonimaging studies would generally be performed as well as radionuclide therapy. Level 3 This level is appropriate for an academic institution where there is a need for a comprehensive clinical nuclear medicine service, human resource development and research program. Radionuclide therapy for inpatients and outpatients is provided.

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Textbook of Radiological Safety IN-VIVO DIAGNOSTIC FACILITY Introduction An in-vivo diagnostic facility need optimal space, equipment and manpower. The design and planning should address many factors including radiation safety. The following are very important on radiation safety point of view: 1. Walls and doors of laboratories should be painted with good quality washable paint; 2. Work table tops should have a smooth laminated finish; 3. Floors should be impervious to liquids; 4. There should be an adequate supply of lead containers and shielding lead bricks; 5. Remote handling devices are desirable; 6. Ventilated fume cupboards are desirable. Equipment While the capacity and quantity of individual pieces of equipment needed depend on the volume of the service, minimum requirements are as follows: 1. A collimated scintillation probe and counting system for uptake measurements of thyroid function and other in vitro and diagnostic studies. 2. An isotope dose calibrator. 3. A portable contamination monitor (acoustic dose-rate meter) and/or a survey meter to monitor beta and gamma contamination. 4. A gamma camera with computer and appropriate clinically proven software. Rectilinear scanners are no longer appropriate. If only one gamma camera is funded, it should have its own computer for static, dynamic and preferably SPECT studies with its various clinically proven acquisition and processing protocols. 5. Provision must be made for a reasonable range of collimators (low energy general purpose, high energy, etc.), including a pinhole collimator. Imaging Rooms Imaging rooms should be at least as large as given in the manufacturer’s recommendations, but preferably larger, to accommodate patients on stretchers. A larger area provides a more pleasant working environment and reduces the risk of radiation to staff. In some hospitals, rooms should have double glazed and insulated windows to avoid the buildup of dust. Tight fitting oversize doors and efficient heating, air conditioning and humidity control units are also required. All rooms should have their own 74 separate power supply and stabilizers and be equipped with hand

Planning of Radiological Facility washbasins with hot and cold running water. An intercom and/or telephone are important for facilitating communication. A typical floor plan of a nuclear medicine facility is shown in the Fig. 4.8 and a typical plan of a radioisotope is shown in Fig. 4.9.

Fig. 4.8: A typical floor plan of a nuclear medicine department

IN-VITRO AND RADIOIMMUNOASSAY (RIA) The design and structure of the building can affect the quality of an RIA centre. Premises should generally provide working conditions that are hygienic and spacious. A patient reception area with a waiting room and an area for taking blood samples should be available. The reception area should be equipped with a couch, resuscitation trolley and other special facilities. It is essential to reserve an area for record keeping and the sorting 75 and labeling of samples.

Textbook of Radiological Safety

Fig. 4.9: Typical plan of a radioisotope laboratory

Floors and bench tops should be smooth and of nonabsorbent material to facilitate cleaning and decontamination in the event of chemical or radioactive spillage. Sinks should be conveniently located at each workbench. A separate washbasin, labelled to this effect, should be reserved for the washing of hands of laboratory personnel, with its use prohibited for any other purpose. The washing-up area for glassware, used RIA tubes and reusable pipette tips should have one or two large sinks and a drying oven. A storage room for buffer chemicals, solvents, test tubes and other consumables is provided. A laboratory preparing its own tracers using imported 125I sodium iodide would need a ‘hot’ laboratory with sufficient space to accommodate a fume cupboard, fraction collector and/or high performance liquid chromatography (HPLC) system, as well as a refrigerator in which to store stock solutions of radioactive material. The areas designated for assays are separated from those reserved for other activities such as patient reception, record keeping and computing. Types of RIA Laboratories

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RIA laboratories are graded on the basis of nature and scope of activity. A Grade 1 laboratory is a basic one using reagents, whether obtained in bulk

Planning of Radiological Facility or as commercial kits, from an outside source, with minimal production of reagents confined to standards and quality control material for the simpler analytes. A Grade 2 laboratory would similarly use primary reagents obtained from elsewhere but in addition produce its own tracers, at least for selected procedures, using 125I produced elsewhere in the country or obtained from abroad. A centre that, in addition to all of the above activities, also produces polyclonal antibodies falls into Grade 3. A Grade 3 laboratory could serve as a national or regional reagent production and distribution centre, or organize and operate an External Quality Assurance Services (EQAS). Finally, monoclonal antibody production centers, if they are also engaged in RIA, are classified as Group 4. Radiopharmacy The layout of the department should enable an orderly flow of work and avoid the unnecessary carriage of radioactive materials within the department. It should be away from gamma cameras, patient waiting areas and offices. It is also important to consider whether there are working areas above or below the radiopharmacy laboratory, in order to avoid unnecessary radiation exposure to people working in those areas. The access to the radiopharmacy should be restricted, and for security reasons, laboratories should be lockable. All surfaces of the radiopharmacy, walls, floors, benches, tables and seats should be smooth, impervious and nonabsorbent, to allow for easy cleaning and decontamination. Floor surfaces and benches should be continuous and coved to the wall to prevent accumulation of dirt or contamination. The radiopharmacy needs to be equipped with at least one isotope calibrator so that all activity can be measured accurately. In addition, a reference source (e.g. 137Cs) will be necessary to ensure continuing reliability of the calibrator. Since, radiopharmacies will be handling unsealed sources of radioactivity, contamination monitors will be required to check for any radioactivity that may have been spilt. Storage areas will be necessary for radioactive materials as well as for nonradioactive components used in radiopharmaceutical preparation. These areas will need suitable shielding and, depending on the type of product being prepared, a refrigerator and freezer may also be required. A typical radiopharmacy layout is shown in Fig. 4.10.

Fig. 4.10: A basic radiopharmacy facility

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Textbook of Radiological Safety RADIONUCLIDE THERAPY Introduction The therapeutic use of radionuclides may be a potential radiation risk for both family members and individuals close to the patient, as well as health workers and the environment. Radionuclides must be used in strict accordance with safety measures and any special instructions, and all precautions must be taken to avoid unnecessary exposure to radiation. The following steps are to be taken before commencing therapy procedures. Licensing The administration of therapeutic doses of radionuclides must be under the responsibility of a physician who is licensed under AERB regulations to administer radioactive materials to humans. Radioactive material for diagnosis or therapy should only be used and stored at medical institutions which possess regulatory license. Technical staff, physicists and nurses may also be subjected to licensing. Facility design and construction When designing therapy units, it is important to bear in mind the following: a. Patients must be housed in a separate room, with dedicated bathroom and toilet. b. Access to the treatment room must be controllable. c. Any required shielding must be designed for the proposed floor plan in the eventuality of pregnant patients in adjacent rooms. It is both easier and less costly to design a unit correctly from the start than to modify it later. Close cooperation between the nuclear medicine staff and architects and builders is vital. If an existing space is to be modified, it may be necessary to determine experimentally the adequacy of walls and floors as radiation shields. If any building work is to be performed, a regular inspection of work in progress is advisable to ensure adherence to agreed plans and specifications. In particular, it should be noted that: i. Brick walls often have inadequate mortar joints, which can be a shielding problem. ii. Flooring should be free of open joints and sealed to the walls, a fact often forgotten by builders. Responsibilities The physician administering the therapeutic radionuclide dose is ultimately responsible for taking every precaution to avoid unnecessary radiation to staff, other patients, visitors and the general public. Before commencing therapy, agreement should be reached on medical and radiation safety 78 protocols.

Planning of Radiological Facility Records A record keeping system must be in place before treatment commences. In addition to normal medical records, a logbook should be kept, listing the patient’s name, the radiopharmaceutical and radioactive quantities administered, and the administration date. Training Radionuclide therapy may involve staff outside the nuclear medicine department, especially nurses and medical staff. A little effort devoted to familiarization and training in the medical and safety aspects of radionuclide therapy can avoid potentially serious problems later. ESTABLISHING A RADIOTHERAPY FACILITY Location Radiotherapy departments are usually located on the periphery of the hospital complex to avoid radiation protection problems arising from therapy rooms being adjacent to high occupancy areas. As pointed out in NCRP 49, operational efficiency, initial cost, as well as provision for future expansion and/or increased workload, should be considered when locating a therapy installation. Proximity to adjunct facilities, ready access for inpatients and outpatients, and consolidation of all therapeutic radiological services, however, may be more important than construction cost. For rooms below ground level, the reduction in shielding costs for floors and outside walls should be weighed against the expense of excavation, watertight sealing and of providing access. Access Access to the room for the delivery and replacement of the treatment unit must be considered. Patients may arrive in wheelchairs or on trolleys or beds. Entrance to the room may be through a shielded door or via a maze. It is necessary to include in the room design an open access conduit for dosimetry equipment cables. This dosimetry duct should always be through a secondary barrier so that the primary beam can never strike it. Ideally it should run at an angle through the barrier to the treatment control area. Also, for security purposes, radiotherapy facilities using radioactive sources should be located in areas where access by members of the public to the rooms where sources are used and stored can be restricted. Further, the proximity of source storage facilities to personnel that may respond in the event of a security breach should also be considered. Room Size The room should be large enough to allow full extension of the couch in 79 any direction, with room for an operator to walk around it. The desirable

Textbook of Radiological Safety size depends upon the type of treatments; for example, a total body irradiation (TBI) procedure will require a larger treatment distance to one wall. For intraoperative procedures (IORT) that require extensive support staff and equipment, the room may need to be larger. The accessory equipment such as electron applicators, breast positioning boards, etc. are usually stored within the room, and should be located to minimize the walking distance for each patient set-up. Maze In order to reduce the radiation dose near the entrance, a restricted access passage way leading to the room may be incorporated in the design. This passage way is termed as the maze. Ideally this should be long in length and small in width. The minimum width may be determined by the dimensions of the treatment unit to be delivered by this route or by access for a hospital bed. A maze ensures the exit of the photon radiation after sufficient scattering. A maze reduces the need for a heavy shielding door. If the length of the maze is sufficient, or if there are enough bends, there may be no need for a radiation protection door at the maze entrance. However, it is recommended that a physical barrier such as a normal door(s) or gate be installed to discourage entry to the maze during patient treatment. Linear accelerators usually require a gate to prohibit entry during treatment times and /or motion detectors to detect unauthorized entry, if a shielded door is not required to reduce dose rates. Another advantage of a maze is a route for ventilation ducts and electrical conduits without compromising the shielding. Doors and Interlocks The treatment room is a controlled area and a barrier be installed at the entrance to the maze or treatment room to restrict access during exposures. If a shielded barrier is required to reduce dose rates, a motorized door may be necessary. A motorized door must have a manual means of opening the door in the event of a power or mechanical failure. There should also be an emergency means by which the motion of the door is stopped. Additionally, any motorized door that is too heavy to be stopped manually should have sensors that stop the motion of the door to prevent injury to personnel and patients. All doors, gates, photoelectric beams and motion detectors must be interlocked to the treatment unit to prevent an exposure if a door is open. The interlock must also ensure that when the door is opened the irradiation will be terminated. The radiation output of the device should not be resumed automatically after the door is closed again. The interlock should be failsafe so that safety is not jeopardized in the event of failure of any one 80 component of the system.

Planning of Radiological Facility In certain centers, it is advised that a door-reset switch be situated near the exit from the treatment room at the position where the person leaving the room has a clear view of the room. Only after activation of the reset switch can the radiation be turned on. If there is a maze, this switch should have a delayed action to allow the person time to leave the room and maze after resetting the switch. This switch should be connected in series with a second switch just outside the door. The same person should operate both switches. In cases where the door is clearly visible from the control panel, closing the door may activate the second switch. Only after activation of both switches can the radiation be turned on (IPEM Report 75). In facilities using radioactive sources, a barrier that restricts access to the treatment room outside normal working hours may also meet certain specifications. An unauthorized access to the source can be detected in a timely fashion. To achieve this, a video camera that provides continuous remote surveillance of the device, a photoelectric beam or motion detector system installed in the maze and/or treatment room, or a door interlock can be incorporated. If these devices indicate the potential presence of an unauthorized person, an alarm should indicate this locally and remotely so that personnel can respond in a timely fashion. These technical measures will be independent of any interlocks that terminate the radiation beam during normal operation because they will not be operational when the treatment unit is powered off outside operational times. Treatment Control Area The treatment control area is where the operators control the machine. This area should be close to the entrance to the treatment bunker so that the operators can view the entrance area. The control area should be sufficiently large to accommodate the treatment unit control console and associated equipment. There may be computer terminals for record and verification, electronic portal imaging, hospital information system and dosimetry equipment, as well as closed circuit TV monitors for patient observation. There should be clear access to any dosimetry ducts. Patient Observation and Communication The operator should be able to visually monitor the patient during treatment with closed circuit TV. Two cameras are recommended. These should be situated 15° off and above the gantry rotation axis for optimum observation of the patient on the treatment couch. The cameras should be located far away from the radiation source, consistent with tele-zoom capabilities, to minimize degradation of the image receptor by scatter radiation. There should also be provision for two way audio communication between the treatment control area and the room. A patient activated alarm may be required for patients unable to give an audible call. 81

Textbook of Radiological Safety Ducts and Shielding Ducts and conduits between the treatment room and the outside must be adequately shielded. This includes ducts for cables necessary to control the treatment unit, heating and ventilation ducts, ducts for physics equipment and other service ducts. It is recommended that ducts should only penetrate the treatment room through secondary barriers. No duct with a diameter greater than 30 mm should penetrate the primary shielding. The ducts should be placed in such a way that radiation passing through them will require the least amount of compensation for the barrier material it displaces. No duct should run orthogonally through a radiation barrier. It could either run at an angle through the barrier or have one or more bends in it so that the total length of the duct is greater than the thickness of the radiation barrier. If required, lead or steel plates are suitable materials to compensate for the displaced shielding. To shield the scattered radiation that passes along the duct, it is better to place the additional shielding outside the treatment room, where the radiation has a lower average energy and therefore, less shielding material is needed as shown in Fig. 4.11. Treatment machine cables are usually run below the floor level under the primary or secondary barriers, before bending up to reach the treatment control area. Provided there are no rooms below, additional shielding is not usually required unless the treatment control area is directly behind a primary barrier, and the cable passes beneath the same primary barrier. Water pipes and narrow electrical conduits are usually placed in groups inside a larger duct. It is recommended that they also should not penetrate through barriers, but follow the maze to exit the treatment room as described above or follow a route beneath the shielding barrier. Heating and ventilation ducts should not penetrate through primary barriers because of their large cross-sectional area, which makes it costly to compensate for the shielding material they displace. If the ducts must pass through a secondary barrier, the cross-section of the duct should have a high aspect ratio to decrease the radiation passing through the duct as a result of multiple scattering interactions with the duct/shielding walls. The axis of the duct and the longer side of the duct cross-section should be as orthogonal as possible to the direction of the leakage radiation from the target towards the duct. The amount of additional shielding required to shield penetrations in shielding walls depends on the energy of the radiation beam, the room layout and the route of the duct(s). The shielding must be evaluated carefully if the ducts must penetrate the primary barrier. The recommended placement of these ducts is above a false ceiling along the path of the maze, to exit the maze at or near the external maze door where the photon and/ or neutron fluence are lowest. For accelerators of energies up to 10 MV, usually no additional shielding around the duct is required. For higher 82 energies, an additional shielding is required. If it is necessary for the ducts

Planning of Radiological Facility

Fig. 4.11: A bend in duck to avoid radiation streaming

to pass through the secondary barrier, they should be placed as high as possible to minimize the scattered radiation to personnel outside the room. Conduit Conduits are required for dosimetry cables, beam data acquisition system control cables, quality assurance (QA) equipment cables, and in vivo dosimetry equipment cables. The conduits are usually PVC pipes of 80 to 100 mm diameter included in the concrete formwork. They should be inclined at an angle (20 to 45 degree, in the vertical and horizontal planes), and penetrate through the secondary barrier but not through the primary barrier. If the openings are at least 300 mm above floor level they are more convenient to use. Ideally, the opening in the treatment control area should be at the counter top level and the opening in the treatment room side should be at a different level but within easy reach. Conduits as described above usually do not need additional shielding unless the barrier is constructed of material with a much higher density than 2350 kgm–3. Room Lighting and Lasers To set up a patient for radiotherapy treatment, the room lights should be dimmable so that the field light of the treatment unit and the alignment lasers can be seen easily. It is useful to be able to control the room lights and lasers from the treatment unit control pendant in the treatment bunker. When the field light is switched on the room lights should dim to a pre-set (but variable) level, and the alignment lasers should also be switched on. Since fluorescent lights do not dim very satisfactorily, it is recommended that incandescent lights be used for the dim level. The main room lighting can be fluorescent lights that extinguish when the field light is turned on and the incandescent room lights are used for the dim level. When the field 83

Textbook of Radiological Safety light is switched off, the main room lighting is switched on and the lasers switched off. The dimmable lights may remain on at all times. Karzmark et al, recommend that if junction boxes or alignment lasers are to be inset in the walls, then the voids need to be backed with 40 mm thick steel plate with a 30 mm margin all around. Depending on the occupancy of the adjacent area, it may be acceptable to have a reduction in the shielding over a small area, especially in a secondary barrier. Four alignment lasers are recommended in total. Three lasers projecting a cross: two aligned with the gantry positions of 90° and 270°, and one mounted in the ceiling directly above the isocentre. The fourth laser should project a sagittal line along the gantry axis. This laser is usually mounted on an angled bracket on the wall opposite the gantry. The laser switching should be controlled from the hand pendant, but it is also useful to be able to switch them off independently for QA tests. Construction Materials To house radiation treatment facilities, concrete will usually be the material of choice since it is the least expensive. However, if space is at a premium it may be necessary to use a higher density building material. Table 4.1 lists a range of typical building materials with their densities. Concrete density will vary according to the aggregate used. Most published data assume a density for concrete of 2350 kgm–3. For therapy installations operating over 500 kV, Compton absorption dominates and the shielding material will absorb the radiation according to the density of material. Table 4.1: Building materials and their densities, (IAEA-47) Building material

Density (kg.m-3) Comment

Concrete Barytes concrete

2350 3400-3500

Iron ore with ferrosilicone 4000-5400 Ledite

3844 and 4613

Clay bricks Breeze blocks

1600 1100-1400

Earth fill

1600

Steel Lead (solid)

7900 11340

⎫ ⎬ ⎭

Will vary with mineral content Most commonly used for dense concrete but expensive Range of densities which depend on proportions of ore mixture to sand Pre-moulded high density interlocking blocks from atomic international, Inc May be used for installations up to 500 kV with supplementary lead or steel shielding May be useful in bunker which is below ground level Normally used as supplementary Shielding on an existing treatment room

Concrete is normally specified by strength, with density being of

84 secondary importance. Strength is increased by increasing the proportion

Planning of Radiological Facility of cement in the mix, while increasing the proportion of aggregate increases density. Increasing the amount of water in the mix will reduce the overall density as air pockets may be left as the mix dries out. To guard against air pockets it is customary to vibrate the concrete mix as it is poured. Each barrier should be formed in one pour to avoid seams between different layers. Preformed concrete blocks only have a density of 2000 kgm–3, but some special dense building bricks are available. Examples of such bricks are barites, or barium and magnetite bricks, which have a density of around 3000 kgm–3. If using dense bricks, it is important to use heavy mortars to avoid shine paths between the bricks. Ordinary sand mortar only has a density of 2000 kgm–3. If space is at a premium, then special high density concretes or high density materials such as steel or lead can be used. Steel plate is often used in existing rooms that need to be upgraded. The steel plate is usually formed in 10 mm thick sheets and fixed one layer at a time to the existing wall, taking care that the fixings do not overlie each other. For therapy installations operating above 10 MV, shielding against neutrons must be considered. Concrete contains a relatively high hydrogen content and is therefore efficient at shielding against fast neutrons. The tenth value layer (TVL) for the primary X-ray beam is approximately double that for the photo-neutrons produced by medical linear accelerators, so any shield designed as a primary barrier against X-rays will be more than adequate against photo neutrons. The fast neutrons are reduced in energy by elastic scattering interactions with hydrogen. After a number of collisions they become slow neutrons, which undergo capture reactions with many materials and penetrating capture gamma rays are emitted. The capture gamma ray spectrum in concrete extends to greater than 8.0 MeV and the average energy is 3.6 MeV. The capture of slow neutrons by hydrogen in concrete results in a pronounced peak in the photon spectrum at 2.21 MeV. Boron and cadmium have large cross-sections for the capture of slow neutrons. Boron is incorporated into polyethylene, which has high hydrogen content to form an efficient neutron shield. Slow neutron capture in the boron results in the production of a low energy gamma ray of 0.473 MeV. A 5% composition by weight of boron in polyethylene is commonly used in neutron shielding doors in treatment rooms. Air Conditioning The treatment room as well as the control room should be air conditioned. The opening for the air conditioners should be provided in the specified outer wall of the treatment room in the case of Cobalt teletherapy room. The lower end of the openings should be located at a minimum height of 2.5 m from the floor level outside and further be covered with a baffle arrangements (4.12). The width of the baffle and the length of its vertical 85 portion should be such that 30 cm wide overlap is available all around the

Textbook of Radiological Safety openings. If split type air conditioners are planned, the baffle arrangements are not necessary. However conduit for passing the AC duct is to be provided in the specified wall at an angle to avoid direct scattered radiation passing through it. Warning Signs and Lights It is recommended that an illuminated warning sign be displayed at the entrance to the maze or treatment Fig. 4.12: Baffle design for air bunker. It should be possible to see a conditioner or exhaust warning sign from any position within the treatment bunker. These signs should be mounted at eye level (1650 mm above finished floor level) and interlocked with the treatment unit control. The illuminated signs may have two or three stages. For a two stage sign, the first stage will be illuminated when there is power to the treatment unit, and the second stage will illuminate when the beam is turned on. For a three stage sign, stage one will be illuminated when there is power to the treatment unit, stage two will light when the treatment unit is programmed to deliver a radiation beam and stage three will illuminate when the beam is turned on. A warning sign should indicate the nature of the hazard. If there are controlled areas with restricted access outside the treatment bunker these should be labeled appropriately. A suitable warning sign such as the redlight must be provided at a conspicuous place outside the room and kept ON when the unit is in use, to prevent entry of persons not connected with the examination or treatment. An appropriate warning placard (Fig. 4.13) must also be posted outside the room entrance or door.

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Fig. 4.13: Radiation warning placard

Planning of Radiological Facility Associated Facility The supporting facility of the radiotherapy department such as simulator room, treatment planning system, mould room, Medical physicists room, radiation oncologists room, examination room, nurses room, and record room etc, should be incorporated in the layout as shown in the Fig. 4.14 A typical lay out of a Tele-Cobalt installation is given Fig. 4.15 A and the cross sectional view is given in 4.15 B. Similary, the model plan of a 6 MV and 15 MV linear accelerator is given in Fig. 4.16 and 4.17 respectively. However, these plans are only models for teaching and training purposes, one has to individually design the facility for the local need, by taking into account all the parameters, including the regulatory concern.

Fig. 4.14: A model layout of radiotherapy department

Additional Installation Requirements 1. Fire protection should be provided. Heat detectors or photoelectric smoke detectors are recommended. 2. Electrical protection as per the local regulation (e.g. earth leakage circuit) must be ensured. 3. Two way patient monitoring intercom system with two video cameras mounted in the treatment room and monitors in the control area in addition an independent intercommunication between treatment room and control desk. The intercom should be voice activated or permanent one.

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Textbook of Radiological Safety Model Plan 1

(A)

(B) AA cross section

Fig. 4.15 A and B: A model tele therapy cobalt installation

88

4. CCTV to monitor the patient and treatment, one fixed and one movable camera. Do not locate in the primary beam.

Planning of Radiological Facility Model Plan 2

BB: cross section Not to scale, all dimensions are in meters, concrete density 2.35 g/cc Area : 11.28 × 10.06 = 113.47 sq m, isocenter height =1.295 m from finished floor. L: Lasers, Sagittal laser height: 2.4-2.6 m AA: AA Cross section, BB: BB Cross section, CC:CC Cross sections.

Fig. 4.16: Model layout of a 6 MV high energy linear accelerator

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Textbook of Radiological Safety Model Plan 3

Not to scale, all dimensions are in meters and concrete density 2.35 g/cc. Area : 12.65 × 10.97 = 138.77 sq m, isocenter height=1.295 m from finished floor. L: Lasers, sagittal laser height: 2.4-2.6 m AA: AA Cross section, BB: BB Cross section,CC:CC Cross sections.

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BB: cross section

Fig. 4.17: Model layout of a 15 MV high energy linear accelerator

Planning of Radiological Facility 5. CCTV monitors may be mounted on or under shelf and must be visible during treatment. 6. Provision of In room monitor to have free view from every side of the couch. Do not locate in the primary beam. 7. Lasers should be fixed at a height of 2.4-2.6 m from finished floor level. Do not locate in the primary beam. 8. Provide battery working emergency light. 9. Provide an outside phone line for remote diagnostics modem. 10. Environmental specifications: Humidity range:40% to 80 % relative humidity, non condensing, Temperature range: 19° to 27° C. 11. The room lights, setup lights, closed circuit television system and In room monitor can be controlled by a single room master switch, often outside the room. The room lights can be on a separate circuit. 12. Set up lights are usually located above and to either side of the longitudinal axis. The operation is independent of the pendant and couch controls. 13. Provide a dimmer switch for set up lights, to adjust the illumination level, so that they are dim enough for clear visibility of the lasers. 14. Provide emergency off switches in room, but do not locate in primary beam. 15. Provide a key switch located at the control console to switch ON/OFF the In room monitor and all monitors and printers at the control console. 16. Provide enough power outlets at the control console and also near the back of the accelerator and the modulator. 17. Provision for wedge tray and compensator tray storage. 18. Provision for block tray storage. 19. Provision for electron applicator storage. 20. Provision for immobilization (acquaplast) systems. BRACHYTHERAPY FACILITY DESIGN Brachytherapy is a radiation treatment with sealed radioactive sources that may be placed within body cavities, within the tissues or very close to the surface to be treated. The duration of the treatment may range from a few minutes for HDR brachytherapy up to several days for LDR interstitial therapy. Many different nuclides are available for clinical use. They may be of low energy requiring minimal shielding or high energy requiring the use of specially designed rooms. LDR brachytherapy is performed either by manually loading sources into applicators that have been positioned in the patient or by remote after loading. The remote after loader stores the sources in a shielded position and, when required, will drive them into the applicators. It will also retract the sources during the treatment, whenever a person needs to attend to the patient and also at the end of the prescribed 91 treatment time.

Textbook of Radiological Safety Rooms used for LDR Brachytherapy may not need special shielding. The layout of the room should allow patients to be nursed safely and also to be used for nonbrachytherapy patients. HDR brachytherapy is only performed with remote after loading units,and requires special facilities. When designing a room for brachytherapy, the following points should be considered: i. Which treatment techniques will the room be used for? ii. What is the likely number of patients per day/week/year? iii. How much radioactivity will be used per treatment/procedure? iv. Which nuclides will be used and what is their energy? v. Where will sources be stored prior to use and after their removal? vi. How will the security performance objectives for brachytherapy be achieved? In brachytherapy, the protection must be sufficient to reduce the primary and scattered radiation to the design limit in all directions since the sources are unshielded in all directions. The dose rate within the room will be much more higher and the room will be designated as controlled area. The dose rate outside the brachytherapy room should be reduced to less than 1 mSv per year. The patient receiving brachytherapy will attenuate the radiation. The extent of the attenuation will depend on the energy of the nuclide in use, the size of the patient and the location of the source(s) within the patient. Since brachytherapy sources are not collimated, the shielding requirements will be based on the transmission of the primary beam through the barriers. If possible the room should be designed so that there is no direct line from the door to the patient’s bed. If there is sufficient space for a maze, a protected room door may be unnecessary, but otherwise a leadlined door will normally be needed. A β,γ monitor which measures the dose rate in the patient area should be clearly visible at the entrance to the controlled area. It is recommended that there be remote viewing of the patient from the nurse’s station by closed circuit TV, together with a two way intercom to reduce the amount of time nursing staff need to spend in the radiation environment. It should be possible to view access to the room from the nurse’s station. LDR and MDR Treatment Rooms For remote after loading systems (either LDR or MDR) the treatment room door will be interlocked to the after loading unit so that the radiation exposure of nursing staff is minimized. Mobile lead shields may be used to reduce radiation dose rates when ideal requirements are not possible. The weight and the need to maintain manoeuvrability of the shield limit the thickness and size of mobile lead shields. Lead shields typically have a thickness of 25 mm and a shielded area of 700 to 1000 mm by 500 to 600 mm. They are usually designed to protect the abdomen of a worker who 92 stands behind them.

Planning of Radiological Facility Some after loading machines allow the treatment of more than one patient at a time so a suite of rooms will be required. Space will be needed for the after loading machine itself and the source transfer tubes. Ideally, the after loading unit will be stored outside the treatment room in a separate closed area. This allows for servicing of the unit when a patient not receiving Brachytherapy occupies the treatment room. HDR Treatment Rooms HDR remote after loading units need special facilities. All the walls, the floor and the ceiling will be primary barriers and must be of adequate thickness to protect the staff, who remain outside the room during the patient treatment. It is advisable to limit the position of the source within the room otherwise all the shielding requirements will need to be determined based on the source being in any position within the room. This may make the barriers unnecessarily thick. HDR sources are usually 192 Ir or 60Co. For both sources, the high activity and HDR require that the room have concrete barriers 400 to 800 mm thick. They will also need a heavy lead door unless a maze has been included in the design. HDR units are often installed in former radiotherapy treatment rooms that already have sufficiently thick walls, ceilings, floors and mazes or shielded doors. In HDR brachytherapy, the patients are often treated directly after the appliances have been positioned. Ideally, there will also be an X-ray facility within the room so that the correct placement of the applicators can be confirmed immediately prior to the treatment being delivered. A waiting

Fig. 4.18: Model plan for a HDR brachytherapy room

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Textbook of Radiological Safety area for a patient on a trolley may be required where the patient may be nursed while the treatment planning calculations are completed. A HDR facility should have an interlocked room door so that the source is returned to the safe position whenever the door is opened, and there should be a radiation warning sign at the room entrance indicating the ‘on-off’ status of the source. A model layout of a HDR room is shown in Fig. 4.18. Area: 9.5 × 6.6 = 62.7 Sq.m. Wall: Concrete 45 cm, density 2.35 g/cc Z.M: Zone monitor at 2 m from floor D: Door ordinary with a glass to peep window. BIBLIOGRAPHY 1. AERB safety code: Brachytherapy sources equipment and installations, AERB / SC / MED-3. 2. AERB safety code: Medical diagnostic X-ray equipment and installations, SC / MED-2 (Rev.1). 3. AERB safety code: Nuclear medicine facilities, SC/MED-4(Rev.1). 4. AERB safety code: Telegamma therapy equipment and installations, SC/ MED-1. 5. Basic radiological physics, Thayalan K. Jaypee brothers Medical publishers (P) Ltd. New Delhi 2001. 6. IAEA safety report series 47: Radiation protection in the design of Radiotherapy facilities, 2006. 7. Nuclear medicine resources book, IAEA, Vienna, 2006. 8. Planning of Teletherapy installations, Users guide, BARC/ RPSD /RASS / TELE-3, 1995.

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Chapter

5

Radiation Monitoring

Radiation exposure must be monitored for both personnel safety and regulatory purposes and it should be carried out periodically. It should also ensure the safety of personnel, patients and the public. The Atomic energy (Radiation protection) rules, 2004 (Earlier RPR-1971, Atomic Energy Act, 1962) insists the radiation monitoring a mandatory one. As per the rules all radiation workers should be monitored with a suitable radiation detecting device. There are two type of monitoring namely (i) Personnel Monitoring and (ii) Area monitoring. PERSONNEL MONITORING The aim of personnel monitoring is stated as follows: (i) Monitor and control individual doses regularly in order to ensure compliance with the stipulated dose limits, (ii) Report and investigate over exposures and recommend necessary remedial measures urgently, (iii) Maintain life time cumulative dose records of the users of the service. Hence, the radiation received by all the radiation workers during their work should be regularly monitored and a complete up to date record of these doses should be maintained. Personnel monitoring is usually done by employing (i) Film badges or (ii) Thermoluminescent dosimeters (TLD) or optically stimulated luminance dosimeter (OSL), and (iii) Pocket dosimeter. The personnel monitoring devices provide (i) occupational absorbed dose information, (ii) assurance that dose limits are not exceeded, and (iii) trends in exposure to serve as check on working practice. In India, country wide personnel monitoring service is offered by the Personnel dosimetry and dose record section, Radiological Physics & Advisory Division (RPAD), CT&CRS building, Bhabha Atomic Research Centre (BARC), Anusaktinagar, MUMBAI-400094.The BARC has accredited M/s Avanttech laboratories (P) Ltd, Chennai and M/s Renantech Laboratories (P) Ltd, Mumbai, to provide personnel monitoring services in India. The requirements of an ideal personnel monitoring systems are (i) instantaneous response, (ii) distinguish between different types of radiation, (iii) accurately measure the dose equivalent from all forms of ionizing radiation with energies from keV-MeV, (iv) independent of angle incidence, (v) small, light weight, rugged, easy to use, (vi) inexpensive, unaffected by environment conditions (heat, humidity, pressure), and (vii) unaffected by

Textbook of Radiological Safety non ionizing radiation. No such dosimeter, satisfying all the above features is available as on date. However, one can be satisfied to some extend by selecting a particular type for a given application. FILM BADGE A film badge is used to measure external individual doses from X, beta, gamma and thermal neutron radiations. It consists of a film pack loaded in a film holder having suitable metallic filters. The film holder is made up of plastic with stainless steel lining as shown in the Fig. 5.1. It is capable of holding one or more photographic films of size 4 cm × 3 cm, wrapped inside by a light tight polythene or paper cover. The metallic filters are fixed on both sides of the holder which help to identify the type and energy of incident radiation. There are three types of holders (i) chest holder, (ii) wrist holder, and (iii) head holder. The film should be loaded in the film holders, so that the flap side of the film pack is always facing the body.

Fig. 5.1: Film badge

The film holder has 6 filters namely open, plastic, cadmium, thin copper, thick copper, and lead. All the filters has 1mm thick except thin copper which is 0.15 mm thick. The filters assess the penetrating power of the radiation and thus permit the energy to be estimated. Thus, it will identify alpha, beta, neutron, low energy X-rays, high energy X-rays and gamma rays, over a range of energies from 10 keV to 2 MeV. There are two films in the badge, one is slow and the other is fast. The slow film is meant for 96 recording high exposure. Film badge is worn compulsorily at chest level. If

Radiation Monitoring a lead apron is used, the film badge is worn under the apron at chest level. The film badge worn at the chest level represents the whole body dose equivalent. The supply of film is for a period of one calendar month (4 weeks). When radiation passes through the filter it causes formation of the latent image in the film. After 4 weeks the film is returned to the agency for dose computation, where the film is processed, the optical densities under different filters are measured by a densitometer. Using standard calibration curves, the dose under each filter is evaluated. A control film is always needed to assess the background level of radiation. Each institution should keep one film, loaded in a chest holder as control. This control badge should be kept in a cool, dry and radiation free area. Monthly dose reports are sent to the individual institutions after processing the film packs. These reports contain monthly doses and up to date cumulative doses of the current year. The doses are reported in mSv and the minimum dose that a film badge can detect is about 0.2 mSv. The advantages of film badge are (i) it is a permanent record (ii) nature of exposure, types of radiation and energy can be evaluated, and (iii) least expensive device. Film badge can be used to measure radiation from 10 mR to 1000 R with a accuracy of ± 10%. The film badges are used only by persons directly working with radiation sources. It is also worth to note that the film badge is used to measure the radiation dose to which the user is exposed. It does not protect the user from the radiation. THERMOLUMINESCENT DOSIMETER The film badge has some disadvantages such as fading at high temperatures and humidity, high sensitivity to light, pressure and chemicals, complex dark room procedure and limited self-life etc. Hence, thermoluminescent dosimeter (TLD) badges are used currently in India instead of film badges. It is based on the phenomenon of thermoluminescence, the emission of light when certain materials are heated after radiation exposure. It is used to measure individual doses from X, beta and gamma radiations. It gives very reliable results since no fading is observed under extreme climatic conditions. The typical TLD badge consists of a plastic cassette in which a nickel coated aluminum (Al) card is placed as shown in the Fig. 5.2. 1. TLD card: The TLD card consists of 3 CaSO4: Dy-teflon disc of 0.8 mm thick and 13.2 mm diameter each, which are mechanically clipped over three symmetrical circular holes each of diameter 12 mm, on a nickel plated aluminum plate (52.5 mm × 29.9 mm × 1 mm). An asymmetric V cut provided at one end of the card ensures a fixed orientation of card in the TLD cassette. The card is enclosed by a paper wrapper in which users personnel data and period of use is written. The thickness of the wrapper (12 mg/cm2) makes the measurements equivalent to 10 mm 97 depth below the skin surface. To protect the TLD discs from mishandling,

Textbook of Radiological Safety the card along with its wrapper is sealed in a thin plastic (polythene) pouch. The pouch also protects the card from radioactive contamination while working with open sources. 2. TLD cassette: TLD cassette is made of high impact plastic. There are three filters in the cassette corresponding to each disc namely, Cu + Al, Perspex and open. When the TLD card is inserted properly in the cassette, the first disc (D1) is sandwiched between a pair of filter combination of 1 mm Al and 0.9 mm Cu (thick:1000 mg/cm2). The copper filter is nearer to the TLD disc and the Al should face the radiation. The second disc (D2) is sandwiched between a pair of 1.5 mm thick plastic filters (180 mg/ cm2). The third disc (D3) is positioned under a circular open window. A clip attachment affixes the badge to the users clothing or to the wrist.

Fig. 5.2: TLD badge Al cards and its holder with filters

The metallic filter is meant for gamma radiation, and the perspex is for beta radiation. The filters are mainly used to make the TLD discs energy independent. When the TLD disc is exposed to radiation, the electrons in the crystal lattice are excited and move from the valency band to conduction band. There they form a trap just below the conduction band. The number of electrons in the trap is proportional to the radiation exposure and thus it stores the absorbed radiation energy in the crystal lattice. After radiation exposure the dose measurements are made by using a TLD reader (Fig. 5.3). The reader has heater, Photo multiplier tube (PMT), amplifier, and a recorder. The TLD disc is placed in the heater cup or planchet, where it is heated for a reproducible heating cycle. While heating, the electron return to their ground state with emission of light. This emitted light is measured by the PMT, which converts light into an electrical current (signal). The PMT signal is then amplified and measured by a recorder. The reader is calibrated in terms of mR or mSv, so that one can get direct dose estimation. The discs are reusable after proper annealing. This badge 98 can cover a wide range of dose from 10 mR to 10,000 R with a accuracy of ± 10%.

Radiation Monitoring Now a days widows based computer controlled TLD readers are available. They are capable of analyzing TLD chips, ribbons, powder, discs, pellets, rods and microcubes. They display digital glow curve and temperature profile. They can handle one or more planchets at a time either with manual drawer or computer controlled drawer function. Programmable annealing oven is also available along with the system. TLD badges do not provide a permanent record and it is available for extremity dosimetry and finger dosimetry (ring). LiF can also be used as TLD phosphor, which has wide dose response, 10 mSv to 1000 Sv. Its effective atomic number is close to that of tissue with an accuracy of ± 2% TLD badges are normally worn at the chest level, that is expected to receive the maximum radiation exposure. Most of the radiation workers used to wear the badge at the waist level which is not correct. During fluoroscopy, it is preferable for the radiologist to wear at the collar level in front of the lead apron to measure the dose to the thyroid and lens of the eye, since most of the body is shielded from the radiation exposure. Pregnant radiation workers should wear a second badge at waist level (under the lead apron) to assess the fetal dose. Additional wrist badge is advised for procedures involving nuclear medicine, brachytherapy source handling and interventional radiology.

Fig. 5.3: TLD reader for dose estimation

Guidelines for Using TLD Badge 1. TLD badges are to be used only by persons directly working in radiation. Administrators, dark room assistant, sweepers etc., need not be provided with TLD badges. 2. TLD badge is used to measure the radiation dose. It does not protect the user from the radiation. 3. The name, personnel number, type of radiation (X or gamma), period of use, location on the body (chest or wrist) etc., should be written legibly in block letters on the front side of the badge.

99

Textbook of Radiological Safety 4. A TLD badge once issued to a person should not be used by any other person. 5. Each institution must keep one TLD card loaded in a chest TLD holder as control, which is required for correct dose evaluation. It should be stored in a radiation free area, where there is no likely hood of any radiation exposure. 6. TLD badge should be worn compulsorily at the chest level. It represents the whole body dose equivalent. If lead apron is used, TLD badge should be worn under the lead apron. 7. While leaving the premises of the institute, workers should deposit their badges in the place where control TLD is kept. 8. A badge with out filter or damaged filter should not be used. It is replaced by a new holder. 9. Every radiation worker must ensure that the badge is not left in the radiation field or near hot plates, ovens, furnaces, burners etc. 10. Every new radiation worker has to fill up the personnel data form, and should be sent to BARC, Mumbai or to the accredited agency. 11. All the used or unused TLD badges should be return, after every service period (quarterly) in one lot so as to reach 10th of next month/quarter. 12. Contact for all correspondence regarding TLD badge service, to the officer in charge, Personnel dosimetry & Dose record section, Radiological physics & Advisory division, Bhabha Atomic Research Centre, CT & CRS Building, Anusakti nagar, Mumbai-400094. Optically Stimulated Luminance Dosimeter Dosimeters using optically stimulated luminance (OSL) is also available now a days alternative to TLD. The principle of OSL is similar to TLD except the heating. Instead of heating, laser is used to stimulate light emission. Crystalline aluminum oxide activated with carbon (Al2O3: C) is commonly used as OSL dosimeter. It has broad base response and capable of detecting low doses as 10 mSv. The OSL dosimeter can be re read several times and it can also differentiate between static and dynamic exposures. POCKET DOSIMETER Film and TLD will not show accumulated exposure immediately. In addition to the regular film badges, the radiation doses received by the radiation worker can be assessed by wearing a pocket dosimeter, which gives instantaneous radiation exposure. This is very useful in nonroutine work, in which the radiation levels vary considerably and may be quite hazardous (cardiac cath lab). The main advantage of pocket dosimeter lies in its ability to provide instant on the spot check of radiation dose received by the personnel. Suitable protective measures can be undertaken 100 immediately to minimize future exposures. The dose can be read off directly by the person during or after any radiation work.

Radiation Monitoring It is an ion chamber with a quartz fiber suspended with in an air filled chamber, on a wire frame as shown in the Fig. 5.4. It has a built-in capacitance which can be charged by an external potential (charger). The positive charge is placed on the wire frame, by means of the charger. The quartz fiber is bent away from the frame due to coulombic repulsion. This can be visible through an optical lens system upon which an exposure scale is superimposed.

Fig. 5.4: Pocket dosimeter

These dosimeters should be fully charged prior to their use so that the initial reading of the dosimeter is set at zero. When exposed to radiation, ion pairs are produced in the air. These ion pairs partially neutralize the positive charge, reducing the coulombic repulsion and allowing the fiber to move. Hence, the quartz fiber move closer to the wire frame, that can be seen as down range excursion of the hair line fiber on the exposure scale (graticule). The movement of the quartz fiber is proportional to the radiation exposure, which is measured in Roentgen (R).The Roentgen is the unit of exposure = 2.58 × 10-4 C/kg. The dose in air can be calculated from the exposure, where 1R exposure is equal to 8.76 mGy (0.876 rad) of air dose. The dosimeter is available in different ranges varying from 0-200 mR, 0500 mR, 0-5R, 0-20R,0-200R and 0-600R for measurement of X and gamma rays. It can detect photon energies from 20 keV-2 MeV. For personnel monitoring, smallest range (0-200 mR) should be employed. A typical commercial chamber with charger is shown in the Fig. 5.5. These dosimeters 101

Textbook of Radiological Safety are available both in analog and digital types. Digital dosimeters use either GM tubes or diodes and solid state electronics. The dose measurement range of pocket dosimeter is 10 μSv to 100 μSv. The accuracy of the pocket dosimeter is about ±10%. Pocket dosimeters are small in size and easy to use and do not provide permanent record. Sudden mechanical shock may result in wrong reading. Hence, these dosimeters should be handled with care so as to indicate reliable reading of the doses received. Now a days digital pocket dosimeters are available with easy display of instant radiation measurements. Presently semiconductor diode based pocket dosimeters with digital display are also available. They have good energy and polar response, with reliable readings, matching to TLD badges. They make loud beep sounds for every 15 to 30 minutes on background. The sound become more frequent as dose rate increases, and becomes continuous sound at high radiation fields. The energy range of these dosimeters are 45 keV to 6 MeV and are available in mR and mSv display.

Fig. 5.5: A commercial pocket dosimeter with charger

PERSONNEL MONITORING SYSTEMS AND FEATURES The common problems associated with personnel monitoring dosimeters includes (i) leaving dosimeters in a radiation field, when not worn, (ii) radionuclide contamination of the dosimeter, (iii) lost or damaged dosimeters, (iv) not wearing the dosimeter while working in radiation. If the body is between the dosimeter and radiation source, the attenuation 102 will cause a significant reduction in exposure. Most of the time the workers

Radiation Monitoring do not remain in fixed geometry, while doing radiation work. As a result the radiation exposure becomes multidirectional and the recorded value is the average exposure for that individual. It is with in ± 10-20% of the individual’s true exposure. The various type of monitoring systems are summarized in Table 5.1. Overexposure If a person receives more than 10 mSv in one quarter, it will be considered as over exposure and the same is reported promptly to the institution and the individual. As per the existing AERB regulatory limits, the effective dose constraint for consecutive 5 years shall be 100 mSv, i.e. average 20 mSv for every year of the sliding 5 years block, the dose limit in any single year should not exceed 30 mSv. The Radiological Physics and Advisory Division (RPAD), BARC will advise the respective institution to take the following actions: 1. The radiological safety officer (RSO) of the concerned institution should examine the working conditions and the circumstances that might have resulted in to the above excessive exposure and report the details to Personnel dosimetry and Dose records section, RPAD, BARC, in the given proforma within 15 days, from the date of receipt. 2. A written statement from the individual, explaining the causes for the reported exposure should also be forwarded along with the RSO investigation report. This is to take preventive steps to avoid such exposure in future. Table 5.1: Comparison of personnel monitoring systems Dosimeter

Radiation

Range

Features

Film badge

β, γ, X-rays

Permanent record

TLD

β, γ, X-rays

γ: 0.1-15,000 mSvb, β: 0.5-10,000/mSv 0.01-106/mSv

OSL

β, γ, X-rays

0.01-106/mSv

Pocket dosimeter

γ, X-rays

0-200 mR 0-500 mR 0-5000 mR

No permanent record Patient dosimetry Reread dosimeter Differentiate static and dynamic exposures Special monitoring Direct reading

AREA MONITORING The assessment of radiation levels at different locations in the vicinity of radiation installation is known as area monitoring or radiation survey. These measurements will give an idea about the radiation status of the installation. On the basis of measurements taken, one could confirm the adequacy or 103

Textbook of Radiological Safety inadequacy of the existing radiation protection status. In case the radiation levels are found to be higher than the permissible levels, suitable remedial measures can be taken. Instruments used for the above purposes are called radiation survey meters and area monitors. In general, any survey meter/ area monitor should consists of two main parts namely: 1. A device which detect the radiation, and 2. A display system to measure the radiation. These instruments differ from each other in the medium in which the response takes place and in the method by which the response is detected and quantified. Following are the different type of meters generally used for radiation survey and area monitoring: 1. Ionization type (air) 2. Geigher-Muller (GM) type (Neon and halogen), and 3. Scintillation detector type [NaI (Tl), ZnS (Ag)]. Selection of a particular detector depends on a variety of factors like type of radiation to be detected and quantity to be measured, response of detector for the energy and type of radiation etc. They can be used as portable radiation survey meters, capable of measuring radiation count rate in mR/h or μR/h. They are available in the form of vehicle mounted radiation meters, zone monitors and door way mounted meters etc. Ionization Chamber Survey Meter Ionization chamber usually consists of an outer cylinder (cathode) coated inside with graphite to make it conducting and a central electrode (anode) insulated from the chamber wall (Fig. 5.6). The cylinder is filled either with air or suitable gas acting as an interacting and detection medium and a suitable voltage is applied across the electrodes. When the chamber is exposed to radiation, the radiation produces ionization in the gas. Under suitable electric field, positive and negative ions are collected respectively by cathode and anode of the chamber. The movement of ions produces an electric current in the outer electronic circuit of the chamber. The strength of this current is proportional to the number of ionization events caused by the energy absorbed in the air chamber and will serve as a measure for quantifying the exposure/exposure rate or dose/dose rate. In ionization chambers the electric field applied is only just sufficient to collect all the primary ions produced by radiation, before they recombine. During this mode, any change in the applied voltage will not affect the number of ionizations produced in the chamber by the radiation. Number of ionizations will be purely dependent on the energy dissipated in the medium. Hence, energy discrimination is possible with ionization chambers for heavily charged particles by pulse height analysis. Since ionization chambers collect only primary ions, electronic amplification of charge is necessary for display purpose. This makes ionization chamber based 104 instruments somewhat delicate and susceptible to extreme climatic

Radiation Monitoring conditions. For X-ray and gamma ray dose measurements, these are operated in current modes. These current is very small (pico-nano Ampere) and requires very sensitive electrometers for measurement. Ionization chambers for low level X-ray monitoring (exposure/ exposure rate) are fabricated out of air-equivalent materials (bakelite, tufnol) and they can be used over a wide range of energies from 7 keV to 2 MeV. A typical survey meter consists of a 500 cc chamber connected to a battery operated electrometer and can measure exposure rates from a few mR/h to about 10 R/h. Some of these are provided with an end window of thin mylar film for beta radiation detection.

Fig. 5.6: Ion chamber

Ion chambers for radiotherapy are fabricated with phenolic wall material with 200-350 cc chamber volume and operated both in dose and dose rate mode. It is recommended to use pressurized ion chambers (8 atmospheres or 125 psi) for in radiotherapy. They provide enhanced sensitivity and improved energy response for the measurement of dose and dose rate. They allow fast response time to radiation leakage, scatter beams and pinholes. In addition, the low noise chamber bias supply provides for fast background settling time. It is capable of measuring gamma energy >25 keV, and beta energy >1 MeV. Ionization chambers are used whenever accurate measurements are required. They approximate the condition under which the roentgen is defined. Ion chambers are used to measure X-ray machine outputs, estimate radiation levels in brachytherapy, and in monitoring radionuclide therapy patients, and survey the radioactive material packages. Ion chambers are influenced by changes in temperature, pressure, photon energy and exposure rate. These limitations are less important in medical applications (5% loss of exposure rate at 10 R/hr). Ion chambers are capable of monitoring higher radiation exposure rate levels, and available in different ranges: 0-5 mR/hr, 0-50 mR/h, 0-500 mR/ h, 0-5 R/h, and 0-50 R/hr. They response slowly (8-2 seconds) to rapidly changing exposure rates and hence needs warm up and stabilization before 105 measurements are made.

Textbook of Radiological Safety Now a days survey meters are provided with lot of special features like auto ranging and auto zeroing, optional beta slide, simultaneous measurement of dose and dose rate, operated by two 9 volts alkaline batteries check source, communications interface with windows based excel add-in for data logging, programable flashing LCD display and audible alarm with dose equivalent energy response (SI units). GM Type Survey Meters In GM survey meter a higher electric field (500-1300 V) is applied between anode and cathode of a chamber, which is filled with a gas of low electron attachment coefficient (e.g argon and neon). The electrons produced in the chamber will have sufficient energy to produce secondary and tertiary ionization during their acceleration towards anode. This results in an amplification of ionization events in the chamber. This is known as gas amplification (avalanche) which depends on the nature of gas and the pressure of gas. In GM counters, use of very thin wire as the anode enables the production of high electric field close to the anode. Also the primary avalanche is followed by a successive avalanches due to secondary phenomena (excitation of gas atoms and production of UV photons). Hence, the whole wire is covered by a sheath of electrons. The gas amplification ( ≈ 108) is independent of primary ionization (i.e. type and energy of radiation). GM type instruments are very sensitive and useful for monitoring of low level radiation. Since electronic amplification is not necessary, the electronic circuit of GM is very simple, compared to that of ionization chamber. This feature makes the GM type instruments rugged and less costly. GM counters used for radiation monitoring generally use a mixture of gases (argon, neon and chlorine/ bromine). It detect the presence and provide a semi quantitative estimate of the radiation field magnitude. It provides measurements in counts per minute (cpm). It also provides an approximate measurement of mR/hr, since it dose not reproduce the conditions under which exposure is defined. But the relationship between cpm and mR/hr is a complicated function of photon energy. GM counters for X-rays and gamma rays monitoring use copper or chromium cathodes for better efficiency (Fig. 5.7). The primary photons interact with the cathode materials to produce secondary electrons. Since, GM meters are pulsed in nature, they should be used only in X-ray units, that emits continuous X-rays. They should not be used in X-ray units, that emits pulsed X-rays (e.g. Linear accelerators). GM type meters mainly used as radioactive contamination monitor with thin window (1.5-2 mg/cm2), and large surface area. It will respond to alpha (> 3 MeV), beta (> 45 keV) X, and Gamma rays (> 6 keV) radiations. GM detector is sensitive to particle radiation, but relatively insensitive to and 106 gamma radiations. It is suitable to measure natural background radiations,

Radiation Monitoring which are 50-100 cpm. It is mainly used in nuclear medicine for low level contamination surveys. GM counters have long dead time (100 μsec) and result in 20% loss at 100,000 cpm measurements. They should not be used in high level radiation fields or when accurate exposure rates are required. Radiation Survey Radiation survey is a procedure in which the exposure rates are measured in and around a radiological equipment by using suitable survey instruments. This is to safety status the quality of the radiological unit. It also ensure that the radiation doses received by the radiation workers are as low as reasonably achievable (ALARA) and they are unlikely to receive doses higher than the maximum permissible limits. No machine should be subjected for patient use (either imaging or treatment), until the radiation survey is carried out. It is to be carried out at time of installation, repeated weekly/quarterly/ annually or after every major repair of the radiation equipment. For example, nuclear medicine require weekly radiation survey, quarterly survey for radiotherapy and annual survey for diagnostic radiology. Radiation survey protocol should be made by the hospital for a specific equipment. The various procedures involved in the radiation survey, for each discipline are explained in the following pages.

Fig. 5.7: GM type radiation survey meter

RADIATION SURVEY IN DIAGNOSTIC RADIOLOGY Introduction The aim of conducting radiological protection survey of a diagnostic installation is to ensure that good quality images are obtained with minimum doses to patients. The surveillance program also fulfill the requirement in respect of filter, collimator, leakage radiation, safe work 107

Textbook of Radiological Safety practices and proper installation planning. Separate protocols of radiation survey should be made available for general X-ray unit, mammography, fluoroscopy and CT scanner. But there are some general requirements of survey for all the above equipments. Radiation protection survey is the evaluation of potential radiation exposure levels at various locations in the installation and the leakage levels incidental to the use of diagnostic equipment under specified conditions. The evaluation includes: (i) inspection of the equipment, (ii) examination of its location with reference to controlled and noncontrolled areas in the immediate environment and (iii) measurement of exposure levels in the environment arising from the operations of the equipment. Inspection of the Equipment 1. 2. 3. 4.

Tube housing leakage Total filtration Tube screen alignment Table top exposure.

General Checks

108

1. Ensure that the X-ray diagnostic equipment is so installed that under no circumstances the X-ray beam is directed towards entrance door, patient’s waiting area, other occupied areas in the immediate vicinity of X-ray room, dark room, film storage, opening in the walls etc. 2. Ensure that control panel is sufficiently shielded with lead lined protective barrier having lead glass windows giving clear view of the rest of the room. 3. Check whether the focus- to- table distance is as per the specification. The X-ray unit should permit a focus–film distance of at least 1.0 meter for all normal radiography and up to 2.0 meter for chest radiography. The focus-to-table top distance should not be less than 30 cm for fluoroscopy units. 4. Check whether the timer of fluoroscopy machine is functioning properly. The maximum range of timer should not exceed 5 minutes. There should be provision for audible signal at the end of the preset time. 5. Check the dark room layout and ensure that the safe light and processing unit are adequate. 6. Ensure that protective devices like lead apron, lead rubber gloves etc., are provided and are in good condition. 7. Ensure that the walls for exhaust/ventilation are provided at least 2 meters above the finished floor level outside and otherwise that the openings are provided with sufficient shielding. 8. Ensure that warning sign (red light and placard) is provided at the entrance of diagnostic room to restrict the entry of public during the operation of diagnostic equipment.

Radiation Monitoring Instruments and Accessories 1. A water phantom (30 × 30 × 30 cm3) to simulate patient scatter condition or a plastic bucket ( ≈ 9 liters) full of water. 2. Ionization chamber type survey meter. 3. Measuring tape. Workload To establish the doses that are likely to be received by the radiation workers and public, it is necessary to know the work load. To calculate the work load, the number of exposures (Nj) of various types (j) per week is noted. The average mAs (Ej) for each such exposure should also be noted. Then the workload can be calculated by using the relation given below. A model workload calculation is given in Table 5.2. n

W=

∑N E j

j=1

60

j

=(

) mA min/wk

Table 5.2: A typical calculation of workload Type of examination Chest Skull Extremities Abdomen

mAs per exam. No. of exams. No. of exams. per day per week 15 40 10 100

25 5 20 10

Total mAs per week

25×5 15 × 25 × 5 =1875 5×5 40 × 5 × 5 =1000 20×5 10 × 20 × 5 =1000 10×5 100 × 10 × 5 =5000

Total workload, mAs per week = 8875 Total workload, mAmin per week = 8875 / 60 min = 147.9

Survey Procedure The sketch of the layout of the installation is drawn and dimensions of the room is measured. The location of the control panel, mobile protective barrier, cassette pass box, doors, windows/ ventilators, passages, dark room and patient waiting areas are indicated in the sketch. A water phantom of not less than 30 × 30 × 30 cm3 dimension is set on the table, to create maximum scatter conditions. The source to image distance (SID) is kept as 100 cm. The collimator is opened for its maximum field size. The machine is operated under maximum kVp and nominal mA settings. The radiation levels are measured at various locations, using a ion chamber type survey meter. These locations include control panel, radiologist position, patient waiting area, doors (both opened and closed position), behind 4 walls of the room, ceiling, below the floor (if the unit is not in the basement) dark room any other location of interest. This is repeated for both vertical and horizontal orientations of the X-ray room. 109

Textbook of Radiological Safety With the knowledge of the workload, the radiation exposure per week, at various locations can be calculated by using the relation: Radiation exposure =

Exposure rate measured ( mR/hr ) × W ( mA min/ wk ) 60 × mA

=(

) mR/wk

The measured readings are tabulated as shown below (Table 5.3). Table 5.3: Radiation survey measurements Locations

Exposure rate level (mR/hr) Beam facing up

Beam facing down

Horizontal beam

Example 1: If the unit is operated for 100 mA and the exposure level measured at entrance door is 180 mR/h. Calculate the radiation level for a total work load of 148 mAmin per week at entrance door level. The weekly Radiation level = (180 mR/100 mA60 min) × 148 mAmin / week = 4.44 mR/wk. Example 2: The instrument reading is 360 mR/h (for tube current 60 mA) at the operator position behind the mobile protective barrier. Calculate the weekly exposure received by the operator(assume work load as 148 mAmin/wk). The weekly exposure =(360 mR/60 mA60 min) × 148 mAmin / week = 14.8 mR/wk. Note: The Permissible dose limit to radiation worker is 20 mSv per year or 0.4 mSv per week or 40 mR/week. Hence, the exposure level at the control panel is within permissible limits. Example 3: The exposure level at the corridor is 30 mR/h (for tube current 60 mA), calculate the weekly exposure to the public (assume the workload as 148 mAmin). The exposure level at the corridor =(30 mR/60 mA60 min) × 148 mAmin / week = 1.23 mR/wk. Note: The permissible dose limits for the general public is 1.0 mSv per year or 0.02 mSv per week or 2 mR/week. Hence, the exposure level in the 110 corridor is well within permissible limits.

Radiation Monitoring RADIATION SURVEY IN NUCLEAR MEDICINE Nuclear medicine radiation survey require the following instruments namely (i) Portable ion chamber survey meter and (ii) GM type contamination monitor. These instruments are kept at good working conditions and needs to be calibrated at regular intervals (once in 3 year). Contamination is the major source of spread of radioactive material in nuclear medicine. Hence, contamination control methods are designed to prevent their spread to personnel and other work areas. Contamination is classified as (i) external contamination and (ii) internal contamination. External contamination is not a series health hazard, but the later gives rise to significant radiation exposures. Hence, internal contamination needs to be prevented by proper radiation survey. The effectiveness of the contamination control is monitored by GM counter survey, at the end of the week, followed by swipe test of the areas. The sketch of the layout of the Nuclear medicine laboratory is drawn and dimensions of the various rooms are marked. The location of the machine room, control panel, source storage, injection room, fume hood, sink, active toilet, patient waiting room, examination room, radioactive waste storage, isolation ward and nurses station etc. are indicated in the sketch. GM counter survey is carried out at the above locations and are recorded in cpm. In addition, ion chamber survey are also carried out in the above locations and recorded in mR/hr. This will identify the areas of high exposure rates especially from radioactive waste and waste storage rooms. Swipe Test Swipe tests are performed by using small pieces of filter paper or cotton at various locations of the nuclear medicine laboratory. Later, these swipes are counted under the NaI (Tl) gamma well counter. Areas that are having twice the background levels are said to be contaminated. Effective decontamination methods are employed to bring back the areas to normal level. Additional swipe tests are performed to confirm the decontamination. Personnel hands, shoes and clothing should be monitored for contamination by the contamination monitor. The accepted level of contamination limits are 0.01 μci per 100 sq.cm, for Tc-99m and 0.0001 μCi per 100 Sq.cm for I131 respectively. Radionuclide Therapy I-131 is commonly used for the treatment of thyroid cancer and hyperthyroidism. Once the patient is administered with I-131, it is excreted in all the body fluids including urine, saliva and perspiration. Hence, exposure rates at 1 m from the patient, bedside, doors and in the adjacent rooms should be measured with ion chamber type surveymeter. The 111

Textbook of Radiological Safety measured levels are posted at the adjacent rooms with suitable instructions to the nursing staff and visitors. The exposure rate measurements are repeated daily, until it comes down to 1.2 GBq (33 mCi) at 1m from the patient. After the patient is discharged, the room is decontaminated, followed by GM counter radiation survey for contamination purpose. Spillage Accidents may happen in nuclear medicine due to radioactive spill, which may be minor or major spill. A minor spill is one in which the activity is less than a mCi. If it is more than a mCi then it is called major spill and the Radiation safety officer (RSO) should be informed. He has to investigate and advise corrective measures. As a first step spills should be contained with absorbent material. The area is isolated and posted with warning signal. Decontamination should be carried out from the perimeter of the spill toward the center to spread the contamination. Decontamination is usually done by absorbing the spill and cleaning the areas with detergent and water. A swipe test and GM survey should follow to ensure decontamination. The protective clothing of the personnel involved in the decontamination procedure should also be surveyed with GM counter. If the spill involves volatile radionuclides, then it may lead to internal contamination, warranting bioassays. In the bioassay the personnel’s thyroid is subjected for external counting with a NaI (Tl) detector for radioiodine. This is followed by radioactivity measurement of urine. RADIATION SURVEY IN RADIOTHERAPY Setting up a Radiotherapy facility involves three major steps namely installation, acceptance testing and commissioning. The vendor does the installation part, while the hospital physicists take care of the acceptance testing and commissioning. After the installation, the medical physicists should carry out radiation protection survey of the installation. The survey will ensure that the exposure levels outside the room will not exceed the permissible limits, considering the dose rate, machine on time, use factors, and occupancy factors for the adjacent areas. A good survey program includes checking the equipment specification, calibration of the machine, measurement of head leakage, area survey, testing of interlocks, warning lights, and emergency switches. The survey should duplicate the conditions that are expected during the patient treatment in terms of work load, use factor and occupancy factor. Radiotherapy uses three category of equipments namely (i) linear accelerator, (ii) Tele-Cobalt unit and (iii) Brachytherapy systems/sources. Hence, specific radiation survey procedure is essential for each one. The following radiation survey meters and items are made available for a 112 successful survey program.

Radiation Monitoring 1. 2. 3. 4. 5.

Pressurized Ion chamber survey meter GM type Contamination monitor Neutron survey meter (BF3) Radiographic film Water phantom.

Linear Accelerator Source ON Position Leakage The gantry of the linear accelerator unit is placed at 180°. Choose 20 measurement points located on the surface of a sphere of radius 2 m from the source. Consider 2 points on the poles of the sphere, 4 equally spaced points on its equator and distribute the remaining points uniformly on the surface of the sphere. Now, the collimator is closed completely and it is covered with 2 TVL of lead shielding (Fig. 5.8). The ion chamber is positioned at any one of the location point at 2 m from the source. The CCTV camera is positioned, to cover the ion chamber location. This will enable us to read the ion chamber through the TV monitor. The machine is switched ON and the ion chamber reading is noted. The ion chamber position is changed to different location points around the head at the same 2 m distance from the source. The readings in exposure rates (mR/h) are noted for all the 20 locations individually. The tolerance limit is 0.2 % of the useful beam dose rate at the treatment distance.

Fig. 5.8: Linear accelerator, radiation survey leakage measurements in mR/h at 2m, for source ON condition

Area Survey The sketch of the linear accelerator installation is drawn on a paper. The occupancy around the installations, controlled area and uncontrolled areas 113

Textbook of Radiological Safety are marked in the drawing. Mark number of locations in the drawing, in which the exposure rate is to be measured. These locations may includes control panel, door all four sides, above ceiling, below floor and patient weighting area etc. The linac machine is set at 100 cm (SSD), with maximum field size (40 cm × 40 cm). The gantry is kept at 0 degree position. A water phantom (30 cm × 30 cm × 30 cm) is kept in the couch to create maximum scatter condition. Now the unit is switched ON and the exposure rate is measured in the above locations by using a wide range ion chamber survey meter. It is repeated to complete the measurements in all the locations. Similarly, the exposure rate measurements are repeated for different gantry positions of 90,180 and 270 degree. The readings are tabulated as shown below (Table 5.4): Table 5.4: Measured exposure rates in mR/h at different locations Survey meter: Wide range ion chamber Make: .......................................................... Model: .......................................................... Gantry position

A

B

C

D

E

Date: .................................. F

G

H

I

0° 90° 180° 270°

Cobalt-Teletherapy Machine Survey In the case of Cobalt-teletherapy machine radiation survey, the head leakage for both source OFF and source ON conditions and area survey of the installation are the essential procedures to be carried out. Head Leakage-source OFF Condition The gantry of the teletherapy unit is placed at 90 or 270 degree and the machine is switched off. Measure the exposure rate (mR/h) at 5 cm from the surface of the source head, at eight different positions around the gantry as shown in the Fig. 5.9. A wide range ion chamber may be used for the measurement. The tolerance limit is < 20 mR/h, when the unit is loaded with maximum capacity source. Similarly, the exposure rates (mR/h) are measured at 1m distance from the source, for different locations (8) around the source. The tolerance limit is < 2 mR/h on the average and 10 mR/h maximum in any direction, at a distance of 1m from the source. Head Leakage Source ON Position

114

The gantry of the machine is positioned at 180 degree and the collimator is closed completely and it is covered with 2 TVL of lead shielding (Fig. 5.10).

Radiation Monitoring

Fig. 5.9: Cobalt teletherapy machine. Radiation survey, leakage exposure measurements in mR/h, for source OFF condition

Fig. 5.10: Cobalt teletherapy machine, Radiation Survey: Exposure rate measured in mR/h for source ON position

The ion chamber is positioned at 1m from the source around the head. The CCTV camera is positioned, to cover the ion chamber location. This will enable us to read the ion chamber through the TV monitor. The machine is switched on and the ion chamber reading is noted. The ion chamber position is changed to different locations around the head at the same 1m distance from the source. The readings in exposure rates (mR/h) are noted for at least 8 locations around the head. The tolerance limit is < 0.1 % of the useful beam dose rate, measured at a distance of 1 m from the source. Area Survey The sketch of the teletherapy installation is drawn on a paper. The occupancy around the installations, controlled area and uncontrolled areas are marked in the drawing. Mark number of locations in the drawing, in which the exposure rate is to be measured. These locations may includes control panel, door, all four sides, above ceiling, below floor and patient weighting area etc. The teletherapy machine is set at 80 cm (SSD), with maximum field size. The gantry is kept at 0 degree position. A water phantom (30 cm × 30 cm × 30 cm) is kept in the couch to create maximum scatter condition. Now the unit is switched ON and the exposure rate is measured by using a wide range ion chamber survey meter. It is repeated to complete the measurements in all the locations. Similarly, the exposure rate measurements are repeated for different gantry positions of 90, 180 and 270 degree. The readings are tabulated as shown in the case of linear accelerator. Primary Barrier Adequacy This survey will revel and ensure the shielding adequacy of the primary barriers. The collimator is set for larger field size (35 cm x 35 cm) and the 115

Textbook of Radiological Safety gantry is kept at 90 degree. Now the beam is focused towards the primary wall 1.The exposure rate behind the barrier is measured by using the ion chamber survey meter. Then the gantry is set at 270 degree and the exposure rate is measured behind the primary wall 2. Similarly, measurements are made by focusing the beam towards ceiling and basement if any. The measurements are recorded as follows (Table 5.5): Table 5.5: Measured exposure rate in mR/h Survey meter: Wide range ion chamber Make: .......................................................... Model: .......................................................... Measurement position

A

B

C

D

Date: .................................. E

F

G

H

Mean

Primary wall 1 Primary wall 2 Above ceiling Below floor

HDR Brachytherapy Survey The high dose rate Brachytherapy equipment require radiation protection survey measurements for both source OFF and ON condition. When the source is in OFF condition, the leakage radiation levels in mR/h is measured both at 5 cm from the surface of the treatment head and at 1 m from the center of the treatment head at various positions as shown in the Fig. 5.11. A wide range survey meter of ion chamber type can be used to perform the measurements. The measured readings are tabulated as shown in the Table 5.6.

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Fig. 5.11: HDR Brachytherapy system: OFF position leakage measurements

Radiation Monitoring Radiation survey measurements are also carried out when the source is in ON condition. Various locations are selected in and around the HDR installation and radiation levels are measured by using the above survey meter, by simulating the HDR treatment with out patient. The readings are tabulated as shown in Table 5.7. CALIBRATION AND MAINTENANCE OF RADIATION MONITORING INSTRUMENTS Radiation monitoring instruments should be kept in good working condition. They should be periodically checked to confirm that reliable readings are indicated. They should also be checked after any servicing or repairs. The simplest method of checking the instrument performance is to use the instrument just after it has been calibrated by the manufacturer and to record for future reference, the exposure rate at a specific distance from a radiation source of known strength. Performance check can then be made at any time, by comparison of the recorded reading with the check reading made at the same distance from the radiation source, after making necessary corrections for radioactive decay of the radiation source. If the check reading after corrections varies considerably, the instrument should be got serviced and recalibrated by the manufacturer. In addition, the operational and handling instructions should be scrupulously observed to ensure prolonged and trouble free performance of the instrument. Table 5.6: Radiation levels at 5 cm and 1 m during source OFF condition Survey meter: ............................................ Make: ............................................ Model: ..................................

Date: .................................. Activity: ...........................

Positions

1 m from the centre of the source housing, μSv/hr

1 . . . 12

5 cm from the source housing, mSv/hr

Positions 13 . . . 16

Table 5.7: Radiation levels during source ON condition Survey meter: ............................................ Make: ............................................ Model: .................................. Position 1 . . 10

Date: .................................. Activity: ...........................

Radiation levels, μSv/hr

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Textbook of Radiological Safety BIBLIOGRAPHY 1. Instructions to High dose rate Brachytherapy users: Nucletron India (P) Ltd, Chennai. 2. Jerrold TB, Seiber JA, Edwin ML, John MB. The essential physics of medical imaging, (2nd edn.) Lippincott Williams & Wilkins 2002. 3. Khan FM. The Physics of Radiation therapy, (3rd edn.). Lippincott Williams & Wilkins 2003. 4. Ramesh C. Nuclear medicine physics, (5th edn.). Lippincott Williams & Wilkins 2004. 5. Thayalan K. Basic radiological physics, Jaypee bothers medical publishers P Ltd, New Delhi 2001.

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Chapter

6

Quality Assurance

INTRODUCTION The term quality assurance (QA) describes a program that is designed to control and maintain the standard of quality set for that program. In medical use of radiation, QA is essentially a set of policies and procedures to maintain the quality of patient care. This is policies only by proper evaluation of the radiological equipment. The general criteria of standard of quality is set by the profession collectively. It is designed specifically for an institution to meet those standards. Professional organizations like American College of Radiology (ACR), and the American Association of Physicists in Medicine (AAPM) have recommended QA programs for Radiological practice. The objective of quality assurance program is a systematic monitoring of the quality and appropriateness of patient care. The QA should be organized as a program which includes the staff training, equipment and facility. The implementation of QA involve administrative, clinical, physical and technical aspects and hence, team work is essential for achieving good quality. The QA programs are developed for a specific application and the following paragraphs will explain the QA procedures related to (i) Diagnostic radiology, (ii) Nuclear medicine, and (iii) Radiotherapy. QUALITY ASSURANCE FOR DIAGNOSTIC RADIOLOGY The goal of QA in diagnostic radiology is to obtain optimal image with minimum radiation dose and at minimum cost. To achieve the above goal, systematic QA programs are to be developed and implemented in all diagnostic X-ray facilities. This will enable us to monitor periodically the performance of the X-ray unit. A Radiological image may be good to look, due to proper density and positioning. If it do not reveal anatomic details, then its image quality is not assured. The radiologists may not be able to extract diagnostic information from the radiograph, which may leads to repeat examination (retake). Retakes result in unnecessary radiation dose to patients, staff and public, increase the workload and cost. The parameters which affects the image quality are applied tube voltage (kVp), tube current (mA), time of exposure (s), focal spot size, contact between film and screen, beam alignment, congruence of optical and radiation fields and Focus to film distance (FFD), film processing conditions and viewing conditions etc.

Textbook of Radiological Safety The quality assurance (QA) program begins with performance evaluation tests of the X-ray unit at the manufacturing site, followed by acceptance tests after the installation is completed. Then QA tests are carried out at regular intervals and also after every major repair. The reasons to test the imaging equipment are to observe the equipment performance at installation in order to determine that it is working properly, to determine that it is currently working as well as it did at the time of installation, or to determine that repairs or modifications have improved recent improper performance. Every QA program involves the following steps that includes (i) performing the QA tests, (ii) record the results, (iii) analyze the result, (iv) take corrective and preventive measures, and (v) repeat the QA again. In general the mechanical characteristics, the control panel display/indicators and the tube housing details are checked initially and it is followed by the set of tests listed below. QUALITY ASSURANCE FOR RADIOGRAPHY UNITS 1. Congruence of Radiation and optical fields 2. Central beam alignment 3. Focal spot size 4. Tube voltage (kVp) 5. Timer check 6. Total filtration 7. Linearity of mA loading stations 8. Timer linearity 9. Out put consistency 10. Tube housing leakage The various tests, their frequency and the tools required for each test are summarized in the Table 6.1. Table 6.1: QA tests, their frequency and tools required

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Test

Frequency

kVp Timer

Once in 3 years Once in 3 months

Test tool

KVp meter Spinning top /KVp/ timer meter and dose measuring meter Out put, mR/mAs Monthly Dose measuring meter Inherent filtration Once in 3 months Dose measuring meter and aluminium absorbers Focal spot size Once in a year Focal spot test tool with nonscreen film cassette Central beam alignment Once in 2 months Beam alignment test tool with screen film cassette Congruence of radiation Once in 2 months Collimator test tool, with and optical field Annually and whenever screen film cassette Grid alignment film density appears Grid alignment test tool nonuniform

Quality Assurance Congruence of Radiation and Optical Fields The optical field in the X-ray equipment is used for defining the radiation field and to limit the same only to the area of clinical interest on the patient. If the optical field and radiation field are not congruent, the area of clinical interest may be missed in the radiograph leading to retake and unnecessary radiation dose to patients. The collimator test-tool is used for testing the congruence of optical and radiation fields. This test tool consists of a fibre glass board of size 24 cm x 27 cm with a rectangular area 20 cm x 16 cm marked on it by coating of X-ray opaque material. This rectangular area is divided into four equal segments by two graduated perpendicular bisectors. Two concentric circles of radii 4 mm and 8 mm are engraved in the centre, which enable the use of this test-tool in conjunction with the beam alignment test-tool. Procedure The table is kept horizontal with the help of a spirit level. A screen type cassette, loaded with a medium speed X-ray film is placed on the table. The collimator test-tool is kept above a screen type cassette. Focus-to-film distance (FFD) is kept as 100 cm, to obtain the shift of fields directly in terms of percentage of FFD. The light field is adjusted to coincide with the rectangular area marked on the test-tool. The film is then exposed under suitable kV and mAs (Fig. 6.1A) and developed. From the radiograph, the shifts (X, X′, Y, Y′) between the edges of optical and radiation fields are measured (Fig. 6.1B). It should be within 2% of FFD. The difference in the dimensions of the optical and radiation fields

(A)

(B)

Figs 6.1A and B: (A) Setting of the Beam alignment and Collimator test tool, (B) Congruence of radiation and optical fields

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Textbook of Radiological Safety (X+X′,Y+Y′) are also recorded. It should be within 3% of FFD. The difference between the sums of the length and width of optical and radiation fields are also computed. The tolerance should not exceed 4% of FFD. Central Beam Alignment If the X-ray beam is not perpendicular to the image receptor, the image may be distorted. This may result in loss of minute details. If grid is used, the distortion will be magnified resulting in total loss of minute details. The beam alignment can be tested using a beam alignment test tool. The test tool consists of a clear transparent acrylic cylinder of inner diameter 6.3 cm, outer diameter 7.5 cm and length 15.2 cm. Acrylic circular discs, each of 6 mm thickness are fastened on both sides of the cylinder. Stainless steel balls of diameter 1.6 mm are co-axially fixed at the centre of both these discs. Beam alignment test is usually carried out along with the test for congruence of optical and radiation field. To do this, the procedure for the congruence of optical and radiation field is repeated. The beam alignment test tool is kept on the collimator test-tool such a way that the stainless steel ball of the lower side of the tool is just above the center of the collimator test-tool. The film is exposed and processed. If the beam alignment is perfect, the image of the top ball will merge with the image of the ball at bottom. The deviation of beam from the perpendicular is determined from the location of the image of the top steel ball in the circles in the radiograph. If the images of the two steel balls overlap, the central ray of the beam is within 0.5° (Fig. 6.2). If the image of the top ball falls within the image of the inner circle, the central ray lies within 1.5 0 from the perpendicular. If it falls between the images of inner and outer circles, the central ray is within 1.50 to 30 from the perpendicular. Tolerance for the beam alignment is 1.50.

(A)

(B)

(C)

Figs 6.2A to C: Interpretation of the image of the two steel balls in the beam alignment test tool

Focal Spot Size

122 The ability for resolving the smallest size of the image (i.e. detail) in a

radiograph depends on the focal spot size. Since, the focal spot size may be

Quality Assurance altered as a result of bombardment of electrons on the target, it has to be checked periodically to ensure that focal spot size is within acceptable limits. Focal spot size is evaluated using the focal spot test tool based on the principle of minimum resolution. Bar/hole test pattern is employed for evaluating focal spot by minimum resolution. At minimum resolution, the edge gradient (penumbra) of one pattern of the pair overlaps with the image of other and the images of both the patterns of the pair cannot be resolved separately. In this condition, the focal spot size (f) is related with line width and magnification (M) as follows: ⎡ M ⎤ f=⎢ ⎥ × line width ⎣ ( M − 1) ⎦ The test-tool consists an acrylic hollow cylinder of about 6.0 cm diameter and 15 cm height. An acrylic circular disc is fastened on one end of the cylinder. Bar patterns engraved on tungsten plate is mounted on this circular disc. The bar test pattern consists of 12 groups of lines (slits) of sizes gradually reducing in dimensions. Each group consists of six lines arranged such that a subgroup of three parallel lines is perpendicular to the other sub-group. The sizes and spacing of the slits in these groups decreases by steps of 16% from 0.84 line pair/ mm to 5.6 line pair /mm. The focal spot size values quoted in the reference Table 6.2, are computed for a magnification of 4/3. This magnification is effected by maintaining the focus to film distance as 60 cm so that the test tool kept on the image receptor will have test pattern at 45 cm from the focus (60 cm -15 cm, which is the length of test-tool). Magnification is calculated as the ratio between focus-to-film distance and focus to object distance (60/45 = 4/3). Table 6.2: Effective focal spot size, for a magnification of 4/3 Smallest group resolved

lp/mm

Effective focal spot size (mm)

1 2 3 4 5 6 7 8 9 10 11 12

0.84 1.00 1.19 1.14 1.68 2.00 2.38 2.83 3.36 4.00 4.76 5.66

4.3 3.7 3.1 2.6 2.2 1.8 1.5 1.3 1.1 0.9 0.8 0.7

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Textbook of Radiological Safety

Fig. 6.3: Bar test pattern for testing focal spot size and its image

To carry out the test, the focal spot test tool is placed on a nonscreen cassette loaded with film. The FFD is kept at 60 cm (Fig. 6.3). Nonscreen technique is necessary to avoid blurring of images of test-pattern. The tool is placed over the cassette so that the vertical patterns are within the anode to cathode direction. The film is exposed and processed. The bar pattern on the radiograph is observed and the smallest group in which all six bars (both vertical and horizontal) are clearly resolved is identified. Minimum resolvable line pair size and the corresponding focal spot size can be obtained from the Table 6.2. The vertical and horizontal groups give vertical and horizontal dimensions of the focal spot. Tube Voltage The applied kilovoltage (kVp) affects the quality and quantity of X-rays reaching the image receptor. This in turn influences the contrast and density of the radiograph. If there is a variation in the the kVp setting, it will affect the image quality. Hence it is necessary to check the kVp settings periodically. This can be done using a kVp meter. The kVp meter, employs two solid-state detectors with different beam hardening filters. When exposed to radiation, the ratio of the signals produced by these detectors will be proportional to the peak tube voltage. A ratio circuit with analog digital circuit (ADC) or a micro processor software system displays the peak kilovoltage, digitally corresponding to a particular ratio of either analog or digital signals. This method is instantaneous and direct reading. Corrections for beam filtration should 124 be applied if necessary.

Quality Assurance The beam centered on the marked area on the top cover of the kVp meter. Proper distance is selected between the focus and the meter. Then it is exposed for a given kVp, mA and time settings. The kVp meter reading is noted. Similar measurements are taken for different kVp settings. The variation between the set kVp and the measured kVp is found. The tolerance is ± 5kVp. Timer Checking If the exposure time set on the diagnostic X-ray unit is not optimal, the radiograph can be under exposed or over exposed. This may leads to repeat examinations. Hence, there is a need to test the timer of the X-ray unit periodically. Manual spinning top (for single phase half wave and full wave rectified systems only) and motorized synchronous tops (for single phase three phase and high frequency systems) can be used to test the accuracy of the timer. Spinning top consists of a rotating circular brass plate with a small rectangular portion cut (hole) at its periphery (Fig. 6.4). Since, the rectangular cut portion is moving with the brass plate, the film receives exposure only when x-ray pulses are produced. Production of X-ray pulses depends upon the rectification of the x-ray unit. A single phase half wave rectified system produces 50 pulses/s and therefore 50 X-ray pulses are generated per second. The time taken for one pulse is 0.02 s (1s/50). If it is a single phase full-wave rectified unit, it will produce 100 X-ray pulses per second and the time taken for each pulse is 0.01 s. Hence, for a set time of 0.5 s the half wave and full wave rectified unit emits 25 and 50 pulses respectively.

Fig.6.4: Spinning top test-tool

To check the timer, the spinning top is placed on a cassette, loaded with film. For a set time, the unit is energized, while the top is rotating. The 125

Textbook of Radiological Safety experiment is repeated to cover the entire range of the timer. The pulses passing through the hole of the circular plate, produces equally spaced rectangular density patterns, on the film. The spacing between the patterns depends upon the speed of rotation of the spinning top. Time =

Number of density patterns on the film pulse frequency

In the case of three phase and high frequency units, synchronous spinning tops are used. These units produce density patterns, which may appear as an arc of continuous trace of density. In such cases, the angle subtended by the arc at the center of the image of the circular plate is measured with a protractor. The exposure time is calculated as the ratio of angle subtended by the arc to the total angle (360°). The speed of rotation (typically one rotation per second) of the disc is suitably selected. Now a days, meters incorporating solid state detectors are available for the measurement of exposure time. Total Filtration All diagnostic X-ray units must have fitted with a minimum thickness of filter, to cut off low energy components from X-ray beam. The low energy X-rays do not contribute to the image formation, but gives unnecessary patient exposure. If the filtration is too high, image contrast will be reduced. Therefore, the total filtration provided for the X-ray tube shall be optimum for patient safety and image quality. For this purpose regulatory bodies recommend total filtration requirement for X-ray machines for different maximum rated tube potentials. Atomic Energy Regulatory Board recommends the total filtration requirements of X-ray diagnostic equipment as follows: Maximum rated tube potential (kVp) Less than 70 70 to and including 100 Above 100

Minimum total filtration (mm Al) 1.5 2.0 2.5

Total filtration includes the inherent filtration and the added filtration. Hence, total filtration evaluation is necessary to verify whether the added filtration is adequate or not. Total filtration of the X-ray tube is evaluated by determining the half value thickness of the beam, by using a pocket dosimeter. The HVT is measured for the maximum operating potential of the tube. The pocket dosimeter is kept at the centre of radiation field of area 20 cm × 20 cm at a distance of 100 cm from the target. For a given kVp and mAs the dosimeter is exposed and the reading is noted. The measurement 126 is repeated and the average is obtained. An aluminum filter of 0.5 mm is

Quality Assurance interposed (at the collimator level) and the measurements are repeated. Similar measurements are repeated for aluminum filters of thickness 1, 1.5, 2, 3, 4 and 5 mm. Transmission curve of the X-ray beam can be plotted on a graph between the absorber thickness and measured dosimeter readings. The absorber thickness for 50 % transmission will be the half value thickness of the X-ray beam. Total aluminum filtration could be determined from HVT using calibration Tables 6.3 and 6.4. Table 6.3: HVT as a function of filtration and tube potential (Single phase generators) Total filtration 30 (mm Al) 0.5 1.0 1.5 2.0 2.5 3.0 3.5

0.36 0.55 0.78 0.92 1.02

Peak potential (kVp) 60 70 80 90 100 Half value thickness (mm Al)

40

50

0.47 0.78 1.04 1.22 1.38 1.49 1.58

0.58 0.95 1.25 1.49 1.69 1.87 2.00

0.67 1.08 1.42 1.70 1.95 2.16 2.34

0.76 1.21 1.59 1.90 2.16 2.40 2.60

0.84 1.33 1.75 2.10 2.37 2.62 2.86

0.92 1.46 1.90 2.28 2.58 2.86 3.12

1.00 1.58 2.08 2.48 2.82 3.12 3.40

110

120

1.08 1.70 2.25 2.70 3.06 3.38 3.68

1.16 1.82 2.42 2.90 3.30 3.65 3.95

Table 6.4: HVT as a function of filtration and tube potential (three phase generators) Total filtration 60 (mm Al)

70

80

2.5 3.0 3.5

2.4 2.6 2.9

2.7 3.0 3.2

2.2 2.3 2.6

Peak potential (kVp) 90 100 110 HVT ( mm Al) 3.1 3.3 3.6

3.3 3.6 3.9

3.6 4.0 4.3

120

130

140

4.0 4.3 4.6

4.6

5.0

Linearity of mA Station The linearity of mA can be tested by measuring the radiation output of the machine. A pocket dosimeter and charger is used to measure the radiation output. The charged pocket dosimeter is kept at the centre of the radiation field of area 20 cm × 20 cm at a distance of 100 cm from the focus. For a fixed kVp and time an available mA station is selected. The tube is energized and the dosimeter reading is noted. The measurements are repeated 5 times, to eliminate statistical variations. Similar measurements are made by keeping the kVp and time constant, for other mA stations. For each measurement X (mR / mAs) is calculated.

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Textbook of Radiological Safety The coefficient of linearity =

X max − X min X max + X min

The coefficient of linearity is evaluated, which should not exceed 0.1. Linearity of Timer To test the linearity of timer, the pocket dosimeter is used. The charged pocket dosimeter is kept at the centre of the radiation field size of 20 cm × 20 cm, at a distance of 100 cm from the focus. The dosimeter is exposed to 50 kV, 200 mA and 0.5 s. The dosimeter reading is noted and the measurements are repeated 5 times. Similar measurements are made for 1s and 1.5 s, by keeping the kVp and mA constant. For each measurement The average and the X (mR / mAs) is calculated. The coefficient of linearity =

X max − X min X max + X min

Coefficient of linearity should not exceed 0.1. Output Consistency To test the out put consistency, the pocket dosimeter is used. The charged pocket dosimeter is kept at the centre of field size 20 cm × 20 cm at a distance of 100 cm from the focus. For a fixed mA and time an available kVp station (say 70) is selected and the tube is energized. The dosimeter reading is noted and the measurements are repeated 5 times. Similar measurements are made for three more kVp, by keeping mA and time constant. For each kVp the average dosimeter reading and the X (mR / mAs) is calculated. The consistency at each kVp station is checked by evaluating the coefficient of variation. 2 ⎛ 1 ⎞ ⎡⎛ ( X − X ) Coefficient of variation (COV) = ⎜ ⎟ ⎢⎜ ∑ i ( n − 1) ⎝ X ⎠ ⎢⎣⎜⎝

1

⎞⎤ 2 ⎟⎥ ⎟⎥ ⎠⎦

Coefficient of variation should not exceed 0.05. Tube Housing Leakage The radiation leakage measurement is carried with an ionization radiation survey meter. For checking the leakage radiation, the collimator of the tube housing is fully closed and the tube is energized at maximum rated tube potential and current at that kVp. The operating time should be greater

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Quality Assurance than the time constant of the survey meter. The exposure rate at one meter from the target is measured at different locations (anode side, cathode side, front back and top) from the tube housing and collimator. From the maximum leakage rate (X, mR/h) for both tube housing and collimator, leakage in 1 hour is computed by assuming workload as 180 mAmin in 1 hour. The maximum radiation leakage at 1 m from the focus, for work load of 180 mAmin in 1 hour is calculated as follows: Maximum leakage = (X mR/hr × 180 mA.min in one hour) / ( 60 min × Applied mA) The tolerance limit of leakage radiation at 1 m from the focus is < 115 mR in one hour. Quality Assurance Test Format 1. Congruence of radiation and optical fields 2. Central beam alignment. Operating parameters: focus to film distance: 100 cm, kV : 50kV, mAs : 20 (mA = ———, s = ——— ) a. Shift in the edges of the radiation field X = .......... cm

% of TFD

X′ = ........ cm

% of TFD

Y = .......... cm

% of TFD

Y′ = .......... cm

% of TFD

Tolerance : 2 % of TFD b. Difference in the dimensions of the radiation and optical fields X + X′ = .......... cm

% of TFD

Y + Y′ = .......... cm

% of TFD

Tolerance : 3 % of TFD c. Difference between sums of lengths and widths of optical and radiation fields X + X′ + Y + Y′ = .......... cm

% of TFD

Tolerance : 4 % of TFD d. Observe the images of the two steel balls on the radiograph and evaluate tilt in the central beam. The tilt in the central beam is —————° Tolerance: Tilt < 1.50

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Textbook of Radiological Safety 3. Focal spot size Operating parameters: Distance: 60 cm, kVp:70, mAs:40-50 (mA= 40-50, s = 1.6) (nonscreen film technique) Large focus size:

Stated —————- mm × ——--mm Measured———--mm × ———mm

Small focus size:

Stated ———-—— mm × ———mm Measured ——-—-mm × ———-mm

Tolerance:

(i) + 0.5 f for f < 0.8 mm (ii) + 0.4 f for 0.8 ≤ f ≤ 1.5 mm

(iii) + 0.3 f for f > 1.5 mm 4. Tube voltage Operating parameters: Distance: 40 – 50 cm Applied kVp

Measured kVp

60 kVp, (40 mAs) 80 kV, (25 mAs) 100 kV, (20 mAs) 120 kV,(15 mAs)

Tolerance : ± 5 kV 5. Timer checking Operating parameters:Distance: 100 cm, kVp: 70, mAs: 40 - 80 Applied time: (i) 0.4 s and (ii) 0.8 s Number of slit patterns on developed film: ———— for (i), time = ———— sec. ———— for (ii), time = ———— sec. Arc measured : ——— for (i), time = ————— sec ————for (ii), time = ————— sec Tolerance : ± 10 % of the set time

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Quality Assurance 6. Total filtration Operating parameters: Focus to detector distance: 100 cm kVp : 100, mAs : 20 (mA : 100 , Time : 0.2 s) Added filter (mm Al )

Output 2

1

Average

Percentage transmission

0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0

Total filtration = —————— mm of Al. Tolerance : 1.5 mm Al for kVp ≤ 70 2.0 mm Al for kVp ≤ 100 2.5 mm Al for kVp > 100 7. Linearity of mA loading station Operating parameters : Distance : 100 cm, kVp : 60, Time : 1.0 s. mA range 1

2

Output 3

4

5

Average mR/mAs (X)

100 200 300

The coefficient of linearity = (COL) Tolerance : < 0.1

X max − X min X max + X min

8. Linearity of timer Operating parameters: Distance : 100 cm, kVp : 50, mA : 200 Time 1

2

Output 3

4

5

Average mR/mAs ( X)

0.5 1.0 1.5

The coefficient of linearity (COL) = Tolerance : COL < 0.1

X max − X min X max + X min

131

Textbook of Radiological Safety 9. Output consistency Operating parameters: Distance : 100 cm Applied mAs kV

Output (mR) 1

2

3

Average mR/mAs (X) 4

5

70 80 100 120 1

⎡ ( X i − X )2 ⎤ 2 ⎢∑ ⎥ n − 1 ⎥⎦ ⎢⎣ Coefficient of variation (COV) = X COV = ———— for ———— 70 kVp, ————— for ————-80 kVp, ———— for ————-100 kVp,———— for —————120 kVp Tolerance : COV < 0.05 10. Tube housing leakage Operating parameters: Applied voltage: ———— kVp, mAs : ————— ( ——— mA, 1.5 s) (Maximum) (minimum) Back X-ray tube Right

Left Collimator Location (at 1.0 m from the focus)

Left

Front

Exposure level (mR/h) Right Front Back

Top

Tube Collimator

Work load = 180 mA min in one hour

Maximum leakage =

132

(X mR / hr × 180 mA min in one hour) (60 min Applied mA)

The tolerance limit of leakage radiation at 1 m from the focus is < 115 mR in one hour.

Quality Assurance QA FOR MAMMOGRAPHY X-RAY UNIT Mammography requires careful optimization of technique and equipment. Even a small change in equipment performance, film processing, patient setup, or film viewing conditions can decrease the sensitivity of mammography. Hence, thorough performance testing is necessary to determine the baseline values and to monitor with periodic quality control testing. The above mentioned tests for general radiography X-ray units can be repeated for mammography also. In addition, the compression needs to be calibrated and the Automatic exposure control device must be checked. Mammography test phantom can be used to optimize the techniques. It simulates the radiographic characteristics of compressed breast tissues, and contains components that model breast disease and cancer in the phantom image. It is composed of an acrylic block, a wax insert, and an acrylic disk attched to the top of the phantom.It is intended to mimic the attenuation characteristics of a standard breast of 4.2 cm compressed breast thickness of 50 % adipose and 50 % glandular tissue composition. The wax insert contains 6 cylindrical nylon fibers of decreasing diameter, 5 simulated calcification groups (Al2O3) of decreasing size, and 5 low contrast disks, of decreasing diameter and thickness, that simulates masses (Fig. 6.5).

Fig. 6.5: Mammography accreditation phantom composed of a wax insert containing 6 nylon fibers, 5 Aluminum oxide speck groups and 5 discs simulating masses

133

Textbook of Radiological Safety Identification of the smallest objects of each type that are visible in the phantom image indicates system performance. As per the recommendation at least 4 fibers, 3 calcification groups , and 3 masses must be clearly visible, at an average glandular dose of less than 3 mGy. The optical density at the centre of the phantom image must be at least 1.2. QA FOR FLUOROSCOPY X-RAY UNIT Table Top Exposure Rate The maximum kVp and mA is selected in the machine. A pocket dosimeter is kept on the table, below the centre of the image intensifier (II) tube field. The unit is energized and the dosimeter reading is noted. Two more readings are taken and average is found in R/min. The tolerance is < 5.7 R/min. Focus to Table Top Distance The distance between the focus to table top is measured in cm. It should not be less than 30 cm. Low Contrast Sensitivity The diameter of the smallest size of the hole, which is clearly seen on the monitor is found. Tolerance limit: A hole of 1/8” diameter. High Contrast Sensitivity Wire mesh lines and bar strips are tested for resolution. The number of mesh lines that are resolved is noted. The tolerance is 30 lines/inch. In the case of bar strips, the resolved bar strips is noted. The tolerance is 1.5 lp/mm. Image Intensifier Assembly Leakage A given kVp, mA and exposure time is selected. The maximum radiation leakage level at 5 cm from the surface of the image intensifier (II) assembly is measured in mR/hr. Then the leakage level for 1 hour is calculated, for a workload of 180 mAmin. The recommended limit for radiation leakage levels at 5 cm from the surface of image intensifier (II) assembly is 100 mR in 1 hour. QUALITY ASSURANCE FOR COMPUTED TOMOGRAPHY Mechanical Tests Alignment of Table Gantry The congruence between the gantry midline and table midline is checked

134 using plumb line. The tolerance should be with in ± 5 mm.

Quality Assurance Scan Localization Light Accuracy A non screen cassette loaded with film is used for this study. The kVp, mAs and slice thickness is suitably selected. The alignment of the internal laser light and the external laser light is checked. The tolerance should be with in ± 2 mm. Gantry Tilt A non screen cassette loaded with film is used for this study. A particular kVp and mAs is selected in the scanner. The gantry is tilted for a set value and the corresponding tilt is measured. The difference between the set and measured gantry tilt is found. The tolerance should be with in ± 3° Table Position /Increment A non screen cassette loaded with film is used for this study. An initial table position is chosen (arbitrary) and a given load is put on the couch. The kVp, mAs and the slice thickness is selected. The table is set with different increments from the reference position; say 1 cm, 2 cm, 3 cm, 4 cm, and 5 cm respectively. At the same time the couch increments are measured correspondingly. The difference between the set couch position and the measured is found. The tolerance should be with in ± 2 mm. Collimator Test Radiation Profile Width A non screen cassette is used for this study. A given kVp and mAs are selected. For a given slice thickness the density profile is recorded on the film. This is repeated for different slice thicknesses. From the density profile, the profile width, i.e., full wave half maximum (FWHM) is found for each slice thickness. The tolerance should be within ± 1 mm. Set slice thickness, mm

Measured density width (FWHM)

X-ray Generator Tests Measurement of Operating Potential The kVp meter is used to do this study. A given mA station is selected on the scanner. Now, a kVp is set on the scanner and the same is measured with kVp meter. This is repeated for different kVp readings. Later, the mA station is changed and the measurements are repeated and the results are 135 recorded as given below. The tolerance should be with in ± 2 kVp.

Textbook of Radiological Safety Set kVp

mA station 1

mA station 2

mA station 3

mA station 4

Measurement of mA Linearity Follow the procedure described for the basic radiography unit, previously. The tolerance limit for coefficient of linearity is ± 0.05 Measurement of Timer Linearity Follow the procedure described for the basic radiography unit, previously. The tolerance limit for coefficient of linearity is ± 0.05 Output Consistency Follow the procedure described for the basic radiography unit, previously. Fix the mAs and slice thickness as constant and vary kVp and make measurements with pocket dosimeter or remote control exposure meter. The tolerance limit for coefficient of linearity is ± 0.05 Resolution Low Contrast Resolution A low contrast resolution test phantom is used for this study. A given kVp, mAs and slice thickness is set on the scanner. The phantom exposed for a given window width. The resolution is measured from the phantom in mm and the percentage of contrast difference is calculated. The tolerance is 5.0 mm at 1% contrast difference(minimum) and the expected is 2.5 mm at 0.5 % contrast difference. High Contrast Resolution A high contrast resolution test phantom is used for this study. A given kVp, mAs and slice thickness is set on the scanner. The phantom exposed for a given window width, using a high resolution algorithm. The size of the smallest resolvable bar /hole pattern is found in mm or lp/cm and the percentage of contrast difference is calculated. The tolerance is 1.6 mm or 3.12 lp/cm at 10% contrast difference and the expected high contrast resolution is 0.8 mm or 6.25 lp/cm. Radiation Dose Tests Measurement of Computed Tomography Dose Index (CTDI) A pencil type ionization chamber with suitable electrometer is used in

136 conjunction with a head/body CT phantom. The phantom is positioned in

Quality Assurance the couch and the ionization chamber is inserted in it. The kVp is set as 80 for a mAs of 100. For a given slice thickness the axial dose and peripheral dose is measured in mGy, then mGy / mAs is arrived. Then, the kVp is changed to 100, and 140 with same mAs and slice thickness and the measurements are repeated. This procedure is completed both for head and body phantoms. The mean of the peripheral dose is found for both head and body phantoms. Then the axial CTDI and the mean peripheral CTDI is found. From the above the weighted CTDI is calculated .The tolerance is ± 20 % of the quoted value (expected) and the minimum is ± 40 % of the quoted value. Measurement

Head phantom

Body phantom

Axial dose, (mGy/mAs) Peripheraldose, (mGy/mAs) (i) (ii) (iii) (iv) Mean peripheral dose CTDTC CTDIP Weighted CTDI, (CTDIW)= 1/3 (CTDTC) +2/3 (CTDIP) = mGy/mAs.

Tube Housing Leakage Repeat the procedure followed previously for general radiography X-ray units. QUALITY ASSURANCE FOR NUCLEAR MEDICINE Quality assurance in Nuclear medicine is essential to ensure that the equipment is always performing to its specifications. The type of QA program and acceptance testing guidelines vary with type of equipment and country. However, the Joint Commission on the Accreditation of Health care Organizations (JCAHO), US and the National Electrical Manufacturers Association (NEMA) guidelines form the basis for most of the QA tests. A typical QA program involve daily measurements of flood-field uniformity, weekly checks of spatial resolution and spatial linearity, and semi annual checks of other performance parameters. All the measurements must be taken under the same conditions (pulse height window width, correction algorithm, and correction circuitry on or off), similar to that of clinical studies. 137

Textbook of Radiological Safety QA FOR GAMMA CAMERA Intrinsic Resolution The intrinsic resolution is determined with out a collimator using a linearity test pattern. The test pattern with strip width of 1 mm is placed on the surface of the NaI (Tl) crystal housing. A point source Tc-99m is placed at distance equal to 5 × UFOV from the camera face. The UFOV is the field of view of the gamma camera after masking off the portion of the camera face affected by edge packing effects. Data are collected, until the peak channel records at least 1000 counts. The count rate should be < 30,000 cps to avoid pile up related mispositioning. Two sets image are taken and recorded, by rotating the text pattern to 90 degree. This will enable to record X and Y resolution. Profiles through the images of the line sources are taken at different locations across the gamma camera face and fitted to a Gaussian function. The FWHM and FWTM (Full width tenth maximum) of the profiles are measured in both X and Y directions. The typical values of intrinsic spatial resolution are 2.5 to 3.5 mm. System Resolution This measurement is made with collimator and should be repeated for each collimator. The source consists of two 1 mm diameter line sources, placed 5 cm apart at a distance of 10 cm from the front face of the collimator. To account scattering, a 10 cm plastic is placed between the sources and collimator and the measurement is taken. Again it is repeated with 5 cm plastic, placed behind the sources. Images are acquired and profiles taken through the image of the line sources are fitted to Gaussian functions, to determine the FWHM and FWTM. The typical resolution is 8-14 mm for Tc-99m. Spatial Linearity Spatial linearity describes lack of spatial distortion. It is a measure of the cameras ability to portray the shapes of objects accurately. This require the slit pattern, line source, and conditions, used for intrinsic resolution. The measurements are taken with two orientations of the test pattern, rotated to 90 degree. This will provide linearity measurements for both X and Y directions. Two more measurements are made from the resulting images. The differential spatial linearity is the deviation of the measured distance between two slits from the actual distance. The maximum deviation of the location of the slits from their true location will give the absolute spatial linearity. Once again it is done for UFOV and CFOV conditions. Uniformity This is studied from flood-field images acquired without collimator. A Tc-

138 99m source is placed at a distance of 5 × UFOV. The counting rate is, 30000

Quality Assurance cps and there should be minimum of 4000 counts in each pixel of the image. Then it smoothed with 9 point (3 × 3) smoothing filter with following weightings: 1 2 1 2 4 2 1 2 1 From this the Integral uniformity and differential uniformity are arrived as follows. Integral uniformity ( % ) =

Max. count Min. count Max. count + Min. count

This is calculated for both UFOV and CFOV. The typical tolerance value is 2 – 4% The differential uniformity ( % ) =

100 × ( high -low )

( high + low )

where high refers the maximum count difference, low refers the minimum count difference for any five consecutive pixels in the image. Counting Rate Performance Two Tc-99m sources are placed about 1.5 m away from the camera face. The total activity is sufficient to cause 20% loss in the observed counting rate, relative to true counting rate. Counting rates are measured with both sources (R12) and with each individual sources, namely R1 and R2. All measurements are taken in the same source geometry. The dead time (t) and 20% count rate loss (R20) are calculated as follows:

⎡ 2R 12 ⎤ ⎡ ( R 1 +R 2 ) ⎤ t=⎢ ⎥ ln ⎢ ⎥ ⎣ R 12 ⎦ ⎣ ( R1 + R 2 ) ⎦ R 20 % =

0.8 ln 0.8 t

Energy Resolution The energy resolution is measured with a flood illumination of the gamma camera face, with out collimator. Tc-99m source is suspended at a distance of 5 × UFOV above the camera face. The resulting pulse height spectrum is analyzed to determine the FWHM of the Tc-99m photo peak. It is usually reported in keV or in % energy resolution based on the photo peak energy. Typical values are 8% to 11% for Tc-99m.

139

Textbook of Radiological Safety System Sensitivity It is to be measured for each collimator. The low energy collimator is tested with Tc-99m, medium energy collimator is tested with In-111 and I-131 is used for high energy collimator. A solution of the radionuclide is placed in 10 cm diameter dish to a depth of 2-3 mm. The source is placed at 10 cm from the camera face. The sensitivity is calculated by drawing a circular region of interest (ROI) around the image of the dish and integrating all the counts in that region. A second image is recorded for an equal imaging time with the source removed, to provide the background.

Sensitivity (cps / Bq ) =

( counts in ROI − background ) time (s) × source activity (Bq )

The typical sensitivity is on the order of 1-1.5 × 10-4 cps/Bq or 0.01% to 0.015 %. QA FOR SINGLE PHOTON EMISSION COMPUTED TOMOGRAPHY (SPECT) Many of the above QA procedures described for gamma camera can also be used for SPECT. However, there are some special requirements as follows: i. Flood field uniformity ii. System alignment. In SPECT, non uniformities can lead to ring (single head system) and arc (multi head system) artifacts. Hence, uniformities of 1% or better are desirable for gamma camera detectors used as SPECT. To detect non uniformities of this order at least 10,000 counts (Poisson statistics) are required per image element. For example, an image matrix of 64 × 64, may require about 41 million counts. It is also necessary to measure the reconstructed image uniformity, which can be done similar to volume sensitivity measurements. In this, an image of a uniform cylinder is reconstructed with suitable attenuation and scatter corrections, by using relevant reconstruction algorithms and filters. The correction tools should not cause any artifacts. The system alignment test verify the mechanical centre of rotation coincidence with the centre of rotation (COR) defined for the projection data (reconstruction). If it is not correct additional blurring or ring artifacts may arise. In the case of multi headed systems, all the heads should be accurately aligned in the axial direction. Otherwise this may enhance the blurring and artifacts. System alignment errors can be measured by recording profiles from different projection angles (0, 90, 180 and 270 degrees) for a point source placed off center in the FOV of the SPECT system. A typical protocol involves N number of projection profiles at equal angle intervals over 360°. 140 For each projection the centroid (rcen, zcen) of the image of the point source

Quality Assurance on the gamma camera face is determined. Where r and z are the radial and axial coordinate respectively. The average COR error and the Axial deviation error are given as follows: N

⎛ 1⎞ Errcor = ⎜ ⎟ ∑ rcen ⎝N⎠ n =1

⎛1⎞ N ErrAX = ⎜ ⎟ ∑ z − z cen ⎝ N ⎠ n =1 where zcen is the point spread function (PSF) centroid in the z direction and z is the mean value of zcen. QUALITY ASSURANCE FOR PET-CT Performance evaluation for Positron emission tomography (PET) [National electrical measurements association (NEMA) NU2-2001 (N-01)] protocol) Spatial Resolution It represents the ability of the system to distinguish between two points of radioactivity in an image. In discrete element systems, smaller detector elements with higher stopping power for 511 keV have the best potential for providing high spatial resolution. In NEMA-01 protocol, it is measured by imaging three 18F point source(PS). A solution of water and 18F with a concentration higher than 185 MBq/ cc is prepared. A drop of the solution was then used to produce three point sources. Glass capillaries with an ID of 1 mm is used to contain the PS. Using a source holder, the three glass capillaries containing the PS are positioned in the the centre of the axial FOV of the scanner at: (i) x = 0 cm, y = 1 cm; (ii) x = 0 cm, y = 10 cm; (iii) x = 10 cm, y = 0 cm. Once in place, the three point sources are aligned (axially) in the scanner FOV using laser lights. Two sets of EM measurements, consisting of 20 acquisitions axially spaced at 0.5 mm are performed in 2D. The two sets of measurements are centred at two axial positions in the scanner FOV: in the centre and at onequarter of the axial FOV (3.8 cm). Acquisition time for each single acquisition was 1 min. The two sets of measurements are then repeated in 3D mode. Image reconstruction of the PS is performed for both 2D and 3D data (FOV 25 cm) and each of the three sources are visualized. Transverse spatial resolution is calculated for each PS position as FWHM and FWTM of the resulting point spread function, by interpolation between adjacent pixels on the radial (vertical) and tangential (horizontal) profiles. An axial profile is derived from the number of counts in each slice against the slice number and axial resolution is measured as the FWHM and FWTM of such a profile. Radial and tangential resolutions (FWHM and FWTM) for each radial 141 position (1 and 10 cm) are averaged over both axial positions.

Textbook of Radiological Safety Sensitivity The sensitivity of a scanner represents its ability to detect annihilation radiation. In the N-01 protocol, the absolute sensitivity of a scanner is measured as the coincidence event rate per unit activity (cps/MBq) from sufficiently low activity line source(LS) suspended within the scanner FOV in the absence of attenuating media. A solution of water and 18F with a concentration greater than 1.7 MBq/ cc is prepared. The LS is prepared by filling a polyethylene tube (ID 1 mm, OD 3 mm) in the central 70 cm and activity is measured. The N-01 sensitivity phantom consists of five concentric aluminium tubes, 700 mm long and stacked one inside the other. The diameters of each tube are: i. 1st tube: ID 3.9 mm, OD 6.4 mm ii. 2nd tube: ID 7.0 mm, OD 9.5 mm iii. 3rd tube: ID 10.2 mm, OD 12.7 mm iv. 4th tube: ID 13.4 mm, OD 15.9 mm v. 5th tube: ID 16.6 mm, OD 19.1 mm. Using a phantom holder, the LS, inserted in the smallest aluminium tube, is centred along the x, y and z axis of the scanner FOV. A set of five 2D EM scans are acquired. In each subsequent scan (60 s each), an additional aluminium tube is added around the LS, so that during the last scan, the LS is surrounded by the 5 aluminium tubes. A second set of measurements (five scans) are taken to estimate the sensitivity in 3D mode. The phantom is then positioned at x=10 cm and y=0 cm with respect to the centre of the scanner FOV and 2D and 3D measurements are carried out following the same protocol as before (for x = 0 and y = 0). Raw data sinograms are used in the analysis of sensitivity. The five scans in each set of measurements are corrected for radioactive decay. The analysis is first performed for each plane. Count rates Rj (j=1,5) are plotted versus the sleeve thickness Xj. The count rate in the absence of attenuation (R0) was calculated by extrapolating the resulting exponential attenuation curve to Xj=0. Rj = R0 exp (– 2 μXj) where µ is the linear attenuation coefficient. The sensitivity for each plane is calculated by dividing the extrapolated R0 by the measured activity. Total system sensitivity is calculated as the sum of sensitivity per plane over the 47 planes. Scatter Fraction and Count Rates The scattering of annihilation photons leads to falsely positioned coincidence events. Variations in design cause PET scanners to have different sensitivities to scattered radiation. The intrinsic scatter function is a measure of the relative system sensitivity to scatter. For a given source 142 distribution, a lower scatter fraction is more desirable, regardless of the

Quality Assurance accuracy of the method for scatter correction, because correction techniques cannot compensate on the noise introduced by the unwanted events and can potentially add bias to the image. The scatter fraction (SF) is defined as the ratio of the scattered events to the total events, which are measured at a sufficiently low counting rate that random coincidences, dead time effects, and pileup are negligible. Total events are the sum of the unscattered events and scattered events. The phantom is a 20 cm diameter solid polyethylene cylinder with an overall length of 70 cm. The phantom has a hole at 4.5 cm from its centre, which goes through the whole phantom, parallel to its central axis. In the hole a Teflon LS (ID 2.3 mm), as long as the phantom, can be inserted to contain radioactivity. The LS is filled in its 70 cm central part with a solution of water and 18F. Two tests are performed, in 2D and 3D, with an initial activity of 2,664 MBq and 1,776 MBq, respectively. In both tests, the phantom is positioned at x=0 cm, y=0 cm and axially centred in the scanner FOV. In both acquisitions, random coincidences are measured by the delayed event (DE) technique. In each test (2D and 3D), 22 EM frames are acquired. Frames 1 to 4 are acquired for 900 s with no delay between consecutive frames. The remaining 18 frames are acquired for 1,500 s with a delay of 1,500s between each consecutive pair of frames. Raw data sinograms are used in the analysis of the scatter fraction and count rates (SF & CR) test. 3D sinograms are rebinned by using a single slice-rebinning algorithm. Scatter component is calculated as for the N-94 over a fixed FOV of 40 cm diameter. Only the final frames of the SF & CR test (when the random rate was negligible, below 1%) are used to calculate the scatter fraction. Scatter fraction is calculated as the ratio between scatter component and total events. The total counts rate within a 24 cm transverse FOV is determined as a function of the radioactivity decay. The T rate (Rtrues) is determined by subtracting the random and scatter rates from the total prompts event rate (Rtotal). The Noise equivalent count (NEC) rate is calculated as follows: NEC =

R trues

R 2 trues + R scatter + kR random

Two NEC rate curves were then generated, with k=1 and k=2. Accuracy of Corrections for Count Losses and Random The same data set as was acquired for the SF & CR test is used. Data are reconstructed with all count rate dependent corrections (dead-time losses and random coincidences) applied. A circular ROI (18 cm diameter) is drawn, centred on the reconstructed images of the phantom. 2D and 3D data sets are reconstructed as for the count rate accuracy test of the N-94. The resulting dead-time and random corrected true image count rate (R)

143

Textbook of Radiological Safety was plotted as a function of activity concentration. The residual dead-time error Δ is calculated as follows: % Δr = 100[1 – (R ÷ Rextrap)] where Rextrap is the linear function of the true count rate extrapolated from the low count rate (where dead-time and random coincidences rate are negligible). Image Quality—Attenuation and Scatter Correction Accuracy It is desirable to compare the image quality of different imaging systems for a standardized imaging situation that stimulates a clinical imaging condition. In this, the scanners image contrast and signal noise ratios are tested under conditions that stimulate a clinical whole body study. Overall image quality (IQ), as well as attenuation and scatter correction accuracy, is evaluated using a phantom simulating a human torso in size and shape. The IQ phantom contains six coaxial isocentre spheres with diameters of 1.0, 1.3, 1.7, 2.2, 2.8 and 3.7 cm. A cylindrical insert of 5 cm diameter, as long as the phantom, is also positioned in the centre of the phantom. The cylinder is a cold insert with a density of 0.30 g/cc to simulate the lungs. Four of the spheres, with diameters of 1.0, 1.3, 1.7 and 2.2 cm, are used to simulate hot lesions, while the other two are used to simulate cold lesions. The phantom is filled with a solution of water and 18F (5.3 kBq/cc), and the spheres with a concentration eight times higher than the background, to simulate a lesion to background (L/B) ratio of 8. In a second experiment, radioactivity concentration in the hot spheres is such that the L/B is 4. Once filled, the phantom is positioned with the spheres both in the transverse plane and along the z-axis of the scanner FOV. To simulate body activity from outside of the scanner FOV, the phantom used for the SF & CR test is positioned at one edge of the IQ phantom. For this test, the LS of the external phantom is filled with an activity of 165.5 MBq. A CT scan of the phantom is used for acquisition (140 kV, 90 mA). For both the experiments (L/B=8 and 4), six interleaved acquisitions (2D and 3D) are performed. The acquisition time for each 2D and 3D measurement is 8 min and 20s and 7 min and 19 s respectively. These times are derived, based on a whole body examination designed to cover a 100 cm axial FOV in 60 min, using a slice overlap for 2D and 3D mode of 5 and 11 slices, respectively. Data were corrected for random coincidences, geometry, normalization, dead-time losses, scatter and attenuation. In order to evaluate the hot and cold sphere contrast, circular ROIs with a diameter equal to the physical size of each sphere are drawn on CT images and copied to PET images. 144 Twelve background ROIs (37 mm diameter) are drawn on the central slice

Quality Assurance and on slices ± 10 mm and ± 20 mm from the central slice. ROIs of smaller size (10, 13, 17, 22, 28 mm) are drawn concentric to the 37 mm background ROIs. Finally, an ROI of 5 cm in diameter is drawn (in each slice of the phantom) on the central cylindrical insert to assess the accuracy of the attenuation and the scatter correction. Different parameters used to evaluate the IQ test are: i. The hot sphere contrast recovery coefficient (HC_RC), ii. The cold sphere contrast (CC), iii. The accuracy of attenuation and scatter correction (ΔA Clung), and iv. The background variability (BVj), are calculated as follows: HC _ RC =

(C (a

) − 1)

hot

/ C bkgd − 1

hot

/ a bkgd

where Chot and Cbkgd are the average of the counts measured in the hot spheres ROI and the average counts in all background ROIs respectively, while ahot/abkgd is the ratio of the activities in the hot sphere and background.

⎞ ⎛C CC = 1 − ⎜ cold ⎟ ⎜ C bkgd ⎟ ⎠ ⎝ where Ccold is the average of the counts measured in the cold spheres ROI. ⎛ Clung ⎞ ΔAClung = 100 ⎜ ⎟ ⎜ C bkgd ⎟ ⎝ ⎠ where Clung is the average counts in the lung insert ROI.

⎛ SD j ⎞ BVj = 100 ⎜ ⎟ ⎜ C bkgdj ⎟ ⎝ ⎠ where SDj is the standard deviation of the background ROI counts for sphere j. Performance Evaluation of CT Performance evaluation tests of CT includes (i) electromechanical, (ii) image quality, and (iii) radiation safety. Detailed information are also available in AAPM report No. 39 (9), AAPM-TG 66 (10) recommendations. Electromechanical Tests These tests include the congruence of gantry laser and imaging plane, localization of CT and pseudo CT centre, orthogonality of table top long axis to imaging plane, accuracy of table vertical and longitudinal movement, radiation and sensitivity profile widths and tests on X-ray generator.

145

Textbook of Radiological Safety Gantry and Couch Congruence of gantry laser with centre of imaging plane and gantry tilt accuracy are verified using ready pack film(Kodak-X-V) adopting the method described in AAPM -39(9). CT center and pseudo CT center an arbitray point exactly 60 cm inferior from CT centre are localized using a commercially available laser calibration phantom. For this purpose 0.1 cm thick tranverse images are acquired at the mid plane of two parallel slabs of the phantom separated by 60 cm. This test is also used to quantify longitudinal table motion accuracy and orthogonality of table top longitudinal axis to the image acquisition plane. Calibrations of table linear scales are verified by moving the table both vertically and longitudinally in steps of 1 cm using an independent measuring scale. Table indexing accuracy and reproducibility are tested by irradiating a ready back film placed perpendicular to the scan plane, under scanner control longitudinal spacing of 0, 5, 10, 20 and 30 cm using 0.1 cm slice thickness. Radiation Sensitivity and Profile Widths A ready pack film placed horizontal to the table top and at the CT centre is exposed using all available slice thickness. The exposed films are measured using film scanner with 0.01 cm step size and the FWHM of optical density profiles corresponding to every slice thickness is obtained. These FWHM values represent the radiation profile widths. Independent verification profile width has to be done by using vendor supplied phantom. X-ray Generator Tests on the X-ray generator include evaluation of peak potential (kVp), timer accuracy(s),mAs linearity and repeoducibility. Non invasive measurement of kVp for different mAs are performed using suitable meters and the method adopted by AAPM-39(9). Linearity of mAs for different kVp is verified by obtaining the product of dose and time at different mA and time settings. IMAGE QUALITY TESTS Image Uniformity and Pixel Noise A 20 cm x 2.5 cm thick water phantom is scanned by using head scanning ptotocol. Mean CT number of water contained with in a circular region of 1 cm2 (400 pixels) is obtained at different locations corresponding to 12, 3, 6 and 9 o’clock positions and at the centre of the phantom using system software. Image homogeneity defined as the edge-to-centre difference in mean CT number is than caculated.Image noise is determined using the 146 relation

Quality Assurance % of noise = (σ × CS × 100) μw where σ is the standard deviation of CT numbers of water within the region of interest, ⎡ ( mm − mw ) ⎤ CS is the contrast scale defined as CS = ⎢ ⎥ , (µm and µw) are ⎣ ( CTm − CTw ) ⎦ the linear attenuation coefficients for the subject material and water respectively, and CTm and CTw are the measured CT numbers of the subject material and water.

Low and High Contrast Resolution Repeat the procedure that are given for QA test for CT scans in diagnostic radiology in this chapter. CT Number Linearity A CT number linearity test phantom is used and the phantom is scanned at 130 kVp with 1 cm slice thickness. The phantom consists of seven cylindrical inserts of different materials (air, perspex, polypropylene, backelite, polystyrene, nylon and Teflon) which stimulate attenuation coefficient of various organs ranging from lung to bone. CT numbers for all these materials are measured from the scan image using system software and compared with the standard value. Radiation Safety The CTDI is measured on the surface and centre for both head and body phantom by using the CT pencil ionization chamber. The scan protocol is usually 1 cm slice thick, axial mode with 80-130 kVp and 100 mAs. Then, the weighted computed tomography dose index (CTDTw) is calculated. For detail procedures refer the QA test for CT scans in diagnostic radiology in this chapter. QA FOR RADIOPHARMACEUTICALS Introduction All radiopharmaceuticals administered to patients must have the safety, quality and efficacy required for their intended use. The employment of short lived radionuclides in radiopharmaceuticals posses problems in quality control testing, since it is not possible to complete the necessary quality control testing before the product’s use-by date. This makes it imperative to employ a range of quick validation techniques in order to test the final product. 147

Textbook of Radiological Safety Control of Starting Materials One of the major aspects of quality control is the source and purity of the non radioactive starting materials. It includes components of kits for technetium radiopharmaceuticals, target materials for use in nuclear reactors or cyclotrons, adsorbents used in columns inside radionuclide generators, and eluents and diluents used in the preparation of the final product. To carry out QA one may require mass spectroscopy and nuclear magnetic resonance spectroscopy etc. which are not usually available in a hospital setup. Hence, it is advisable to purchase materials from radiopharmaceutical manufacturers who might have performed the above quality control procedures on the materials they are supplying. Testing the synthesis of non radioactive materials may require the use of analytical techniques such as infrared and ultraviolet spectroscopy, mass spectroscopy and nuclear magnetic resonance, and the department should ensure that it has access to such facilities. Details of the synthesis and analysis of certain kits are provided in IAEA-TECDOCs 649 and 805. Information on the specifications that radiopharmaceuticals should meet is also available in national and international pharmacopoeias. If the product has been approved for marketing by an appropriate authority, the user department should have little or no testing to perform on it. Continued satisfactory use of the product enables the user to build up confidence in the quality of the supplier. Radionuclide Activity It is necessary to ensure that the correct activity is administered to the patient. Accurate measurement must be taken place during the preparation of radiopharmaceuticals and the dispensing of individual doses. There is therefore a requirement for control of the dose calibrator to ensure its correct functioning and accuracy. It is always advisable to measure the vial before and after dispensing the radiopharmaceutical into the syringe. The difference between the readings gives a more reliable indication of the dispensed activity. Radionuclide Purity Radionuclidic purity is defined as the percentage of the activity of the radionuclide concerned to the total activity of the sample. All radioactive materials are likely to have some radionuclidic impurities, albeit at very low levels, which can make their determination difficult. The situation most relevant to hospitals and clinics is the determination of levels of 99Mo in 99m Tc eluted from a generator. Fortunately, this can readily be determined by a screening method since the principal gamma energy of 99Mo (740 keV) is much higher than that of 99mTc (140 keV). The total activity of a sample is 148 measured in the normal way in a dose calibrator. The sample is then placed

Quality Assurance inside a lead pot 6 mm in thickness, which attenuates virtually all the 140 keV gamma rays of technetium but only approximately 50% of the 740 keV gamma rays of 99Mo, and the activity is remeasured using calibration factors supplied with the instrument. It is then possible to calculate the amount of 99 Mo present and express this as a percentage of the 99mTc. Most pharmacopoeias have a limit of 0.1% of Mo at the time of administration, and any eluates that exceed this limit must not be used. The determination should therefore be carried out on the first eluate of a generator and on other eluates as deemed necessary. Radiochemical Purity The radiochemical purity is defined as the proportion of the total radioactivity of the nuclide concerned present in the stated chemical form. For many radiopharmaceuticals the radiochemical purity will be expected to be greater than 95%, but this is not universally so. Manufacturers will normally declare the radiochemical purity, for which further testing is not necessary. For materials prepared in-house, either totally from original materials or purchased kits, radiochemical purity determinations are useful to establish the suitability of the final product. Low radiochemical purities may lead to an unintended biodistribution of the radiopharmaceutical. For diagnostic agents, this may lead to confusion in the diagnosis and for therapeutic radiopharmaceuticals it can produce significant dosimetric problems. A range of techniques is available for such determinations, but the techniques must be reliable and simple, and preferably rapid, to perform such that, in an ideal situation, the radiochemical purity of materials containing short lived radionuclides can be established prior to their administration. The simplest and most widely used technique is that of planar chromatography, using suitable stationary phases (e.g. paper or thin layers of silica gel) and readily available mobile phases (e.g. saline, acetone and butanone). The choice of stationary and mobile phases is determined by the nature of the radiopharmaceutical, and must be such that the various radiochemical species have different mobility’s. Suitable systems for a range of radiopharmaceuticals are given in IAEA-TECDOCs 649 and 805. The techniques can be carried out with very simple apparatus, for example with beakers or measuring cylinders as chromatography tanks; in view of the scale of the operation only small volumes of solvent are needed. The levels of each species can be determined by scanning the stationary phase with a suitable detector or cutting it into sections and placing each in a counter. However, the limitations of these simple systems need to be borne in mind, since in many of them only certain impurities (e.g. pertechnetate in Tc radiopharmaceuticals) migrate with the solvent. Most of the activity may remain at the point of application on the chromatography strip and 149 thus be unresolved. Alternative techniques such as electrophoresis or HPLC

Textbook of Radiological Safety offer advantages in that they can give more precise information about the radiochemical nature of the species present. Commercial manufacturers of radiopharmaceuticals use HPLC routinely. The technique utilizes the separating power of adsorbent materials packed into stainless steel columns through which a solvent is pumped at high pressure. Different radiochemical species are identified by monitoring the eluate from the column and noting the time at which radioactivity is detected. This technique has limitations in that the apparatus is expensive and may not be routinely available to hospital radiopharmacies. In addition, certain radiochemical species, for example, hydrolyzed reduced Tc in Tc radiopharmaceuticals, may be retained on the column used to achieve the separation and may not therefore be accounted for in the analysis. Recent developments have included the introduction of cartridges containing the same absorbents used in HPLC, but which can be loaded and eluted with syringes. By using appropriate eluents, different species can be selectively removed from the cartridge and, providing a sufficiently high radioactive concentration is used, activity can be determined with a dose calibrator or other simple scalar. Thus, the hospital radiopharmacy can benefit from the resolving power of absorbents used in HPLC, but without the expense of the equipment required. Chemical Purity In addition to the problems of ensuring the correct chemical purity of starting materials for radiopharmaceuticals, there are certain situations where the chemical purity of the final material can be affected by the process used in the preparation. The most likely situation to be met in radiopharmacies is the presence of Al ions in Tc radiopharmaceuticals. These can arise from alumina being washed off the columns used in Tc generators. Very high levels of Al can be toxic to patients, but it is unlikely that such problems will arise from administration of a radiopharmaceutical. However, lower levels can adversely affect radiopharmaceutical formation or stability, for example of colloidal radiopharmaceuticals, where the trivalent Al cation can alter the surface charge of particles and lead to aggregation and hence an altered biodistribution. Aluminium can be detected by a simple colorimetric limit test, using either a solution or indicator strips containing an Al sensitive marker such as chromazurol-S. By comparing the colour obtained with a small volume of the eluate of a Tc generator and that from a solution containing a specified concentration of Al ions (generally 5 or 10 parts per million), it can be determined that the Al content of the eluate is below the specified level and hence suitable for use. Metal impurities may reduce the efficiency of 111–In radiolabelling.

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Quality Assurance Determination of Particle Size Lung imaging agents are normally based on macro aggregates of human albumin. A particle size range of 10-100 mm is generally specified as being optimal. Some pharmacopoeias state that there should be no particles larger than 150 mm. Particle size can be determined by light microscopy, using a graduated slide to ensure that there are no oversize particles and that a suitable range of sizes is present. The limitations of the method are that it is usually only possible to observe a limited number of particles and that prolonged observation subjects the eyes to an increased radiation burden. These limitations can be overcome by reconstituting a macro aggregate kit with saline and observing nonradioactive particles. Colloidal particles cannot be visualized by normal light microscopy and, in situations where it is important to know the particle size distribution, more elaborate techniques such as light scattering or membrane filtration will have to be used. These may not be readily available in hospital radiopharmacies. Particulate Contamination Products for parenteral administration should be free from gross particulate contamination. The use of clean glassware, kits, reagents and equipment is the best way to minimize contamination. However, on occasions, particles can be present in the final solution as a result of coring of the rubber stopper if it is repeatedly punctured. Control can be exercised by visual inspection of the final radiopharmaceutical, while ensuring that adequate measures are taken to protect the eyes. The required level of protection can be achieved by viewing through lead glass screens or by using mirrors to view vials placed behind lead shields. It should be pointed out that such techniques may not detect small amounts of particulate contamination and are not suitable for radiopharmaceuticals which themselves are particulate. Control of pH For some radiopharmaceuticals, control of pH is essential to ensure they retain their original specification. For example, indium (111In) chloride must be maintained at a pH of 1.5. If the pH rises, the material becomes colloidal and unsuitable for labeling reactions. With Tc compounds, the chemical composition and hence biodistribution of DMSA complexes is affected by the final pH of the solution. The normal renal imaging agent must be maintained at a pH below 3.5. The easiest method of determining pH is to use narrow range pH papers, since only small samples are needed. Papers are readily available from a variety of sources. Assessment of pH is subjective and such papers are normally only accurate to about 0.5 of a pH unit. For the majority of radiopharmaceuticals these limitations are not normally detrimental.

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Textbook of Radiological Safety Sterility and Apyrogenicity Radiopharmaceuticals administered parenterally need to be sterile and apyrogenic. Although, these objectives can be achieved by the use of a suitable sterilization technique during preparation of the radiopharmaceutical, it is often necessary to use an aseptic technique to prepare the final radiopharmaceutical, having started with sterile materials (e.g. kits and generator eluate). Control of the environment in which such manipulations take place is important. Sterility testing of radiopharmaceuticals present difficulties and it is often impracticable to apply tests described in pharmacopoeias; this is not only because of the radioactive nature of the material but also, as is the case with Tc radiopharmaceuticals, because the batch may consist of a single container. This introduces serious problems with sample sizes and makes the test statistically unsatisfactory. In addition, there is evidence that microorganisms do not survive in Tc radiopharmaceuticals, and hence allowing them to decay in order to make testing easier can reduce the value of the test. As a compromise it is probably better to withdraw a small sample of the radiopharmaceutical whilst it is still active and place it in a suitable culture medium that can be shielded until decay has occurred. It can then be incubated in the normal way. Alternatively, for Tc radiopharmaceuticals, the culture medium can be added to the remnants of the kit vial at the end of the working day. The vial is kept shielded until inactive and then incubated. Inevitably this means that the result of the test is only obtained retrospectively. In view of these limitations, a more satisfactory technique to ensure sterility of aseptically prepared radiopharmaceuticals involves staff simulating exactly the preparation techniques using culture media. Such tests have the advantages of being more sensitive and of using non-radioactive materials, and can be performed earlier. Determination of the apyrogenicity of injections is currently required only when the volume administered exceeds 15 ml. This rarely occurs with radiopharmaceuticals and hence the test is not usually performed in hospital radiopharmacies. If a hospital is involved in the development of new agents, it may be prudent to assess the apyrogenicity, particularly if materials of animal origin are used in the preparation. The use of the limulus lysate test for pyrogens is now becoming widely accepted in preference to the rabbit test, but rigorous controls must be used to validate the test. Commercial manufacturers frequently use the limulus lysate test in the control of their materials. Ongoing Evaluation of Product Performance Diagnostic radiopharmaceuticals of appropriate quality should have a 152 defined biodistribution within patients. If such observations are made

Quality Assurance regularly, confidence in the quality of the materials being administered to patients is gained. When nuclear medicine images are reported, unexpected biodistributions are sometimes observed and may result from problems with the radiopharmaceutical, or alternatively may be due to the patient’s condition or even the medication the patient may be taking. It is worth trying to determine the cause of the problem. If the problem has occurred with all patients who received that particular batch of radiopharmaceutical, the problem is likely to lie with the product. An example is the visualization of the stomach in patients undergoing bone imaging with a technetium phosphonate complex. This indicates the presence of pertechnetate in the radiopharmaceutical and may have arisen as a result of an incomplete reaction when preparing the kit or of instability after preparation. If this occurs on a regular basis with different batches of the same radiopharmaceutical, action is necessary to eradicate the problem. This may involve review of the methods used in preparation or a change in purchasing patterns of materials. However, it is not acceptable merely to rely on the biodistribution in patients as the only quality control testing to be performed. In situations where an unexpected biodistribution is seen in one patient but not in others who received the same product, a patient related cause might be responsible. If this can be identified, it can provide useful information for future reference and to prevent misdiagnosis occurring. On rare occasions, an adverse reaction may occur in a patient to whom a radiopharmaceutical has been administered. This does not mean that the product is necessarily defective. The prevalence of such reactions has been estimated as 3 per 105 administrations and, as such, departments might not encounter a similar situation for many years. Fortunately, adverse reactions that do occur are generally mild and self-limiting and do not require extensive treatment. The adverse reaction most commonly encountered involves the development of skin rashes a few hours after administration of 99mTc bone imaging agents. Histamine release in the patient is frequently implicated as the cause of the problem, and hence symptomatic treatment with an antihistamine is sometimes beneficial. There are occasions when a severe anaphylactic reaction can occur immediately after administration and prompt action, including administration of adrenalin, may be necessary. Since the occurrence of such events is so low, they should be reported to the manufacturer of the product and, as necessary, to national authorities. In this way a database on the possible reactions that can occur is developed and information can be disseminated. Departments can then be prepared to deal with such events if they occur, thereby enhancing the quality of patient care. SUMMARY Each department needs to have its own quality assurance program to ensure 153 that the products administered to patients are of the desired quality. This

Textbook of Radiological Safety requires the development of appropriate documentation systems, record keeping and quality control testing protocols. These will be influenced by the range of products prepared, the source of the starting materials (e.g. from a commercial manufacturer or prepared in-house) and the facilities used for the preparation. In addition, it is important that the results obtained are reviewed and acted upon where necessary in order to maintain the quality of the products. One vital component in the assurance of quality of products is to have well trained competent staff who have the necessary skills and knowledge to deal with radioactive pharmaceutical products. QUALITY ASSURANCE FOR RADIOTHERAPY Quality assurance (QA) is the method of subjecting the newly installed equipment to an exhaustive performance testing to determine that the equipment is meeting the vendor’s technical specifications and hospital’s clinical specifications. Usually the QA tests are performed at the time of installation, known as acceptance testing and repeated periodically. The periodic QA will ensure the integrity of its basic physical and functional specification through time. This will also ensure that the fundamental parameters have not changed since last measured. Some of the basic QA parameters and procedures are given below. However, individual hospital has to devise their own QA methodology and periodicity to suit their machine and model. QA FOR LINEAR ACCELERATOR Radiation Survey Radiation protection survey involves the measurement of head leakage, area survey and test of interlocks, warning lights and emergency lights. The survey is evaluated on the basis of clinical use, by taking into account the workload, use factor and occupancy factors. The detail procedure of survey is explained in chapter five under area survey. Jaw Symmetry To study jaw symmetry, a machinist’s dial indicator is used. First the gantry is set at horizontal and jaws open to a large field size. The feeler of the dial indicator is made to touch the face of the one of the jaws and the indicator reading is noted. Now the collimator is rotated to 180 degrees and the feeler is touching the opposite jaw and the dial reading is again noted. The difference between the two readings is noted. The symmetry error is ½ of the difference in readings. The procedure is repeated for the second jaw. Spirit level is used to check the collimator angle. The tolerance for symmetry error is 1 mm.

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Quality Assurance Coincidence Collimator Axis, Light Beam Axis and Cross-hairs Coincidence Gantry is set at vertical and the SSD is set at 100 cm. A graph paper is fixed on the couch and the field size is kept as 10 × 10 cm. Switch on the light field and mark the edges of the light field, intersection of diagonals and the position of the cross hairs images. Rotate the collimator through 180 degrees and mark the above parameters in the graph paper. Check the coincidence of light field edges, intersection of diagonals and position of cross hair images. If there is a misalignment it should be adjusted to bring down to coincidence. Optical and Radiation Beam Congruence A therapy verification film back is placed on the couch with SSD of 100 cm. A field size of 10 cm x 10 cm is set and collimator angle is set to 0 degree. The light beam is made on and the light field edges and the centre is marked with lead wires or radio opaque markers. A plastic (2-5 mm) sheet is placed over the film pack to give electronic buildup and eliminate electron contamination. The film is exposed so that a optical density of around 1 is achieved. The procedure is repeated for a collimator angle 90, 180 and 270 degrees. The coincidence of the optical and radiation beam is checked visually or by cross beam optical density profiles (Fig. 6.6). The tolerance is ± 3 mm. Mechanical Isocenter Collimator Rotation A graph sheet is fixed on the couch and front pointer is put on the accessory mount with a gantry angle of 0 degree. With the pointer extended, the SAD is set to 100 cm. Now the pointer tip position is marked on the graph sheet. The collimator is rotated to 90,180 and 270 degree and each time the pointer tip position is noted. A sharp edge or wiggler may be attached to the end of the pointer rod to have effective observation. The tolerance of the isocenter is ± 2 mm diameter. Gantry Rotation In the above procedure, another horizontal rod with fine pointer is positioned by means of a stand. The stand should be kept away to avoid gantry collision. The horizontal rod tip and the front pointer tip are made to coincidence at 100 cm SAD with gantry position of 0 degree. The gantry is rotated to 90,180 and 270 degrees and the displacement between front pointer tip and the horizontal rod tip is observed. The tolerance of the isocenter is ± 1mm.

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Textbook of Radiological Safety

Fig. 6.6: Optical and radiation field field congruence: 9 MeV electron and 6 MV photon beams (For color version see plate 1)

Radiation Isocenter Collimator The gantry is set to vertical, 0 degree with a SAD of 100 cm. A ready pack film is kept flat on the couch. The upper jaws are fully opened and the lower jaws are closed to have a narrow slit of beam. A build up sheet is placed over the film and it is exposed to create a density of about 1. The collimator is rotated to different angles (4-8 angles) and each time the film is exposed. The procedure is repeated for upper jaws of narrow slit, while the lower jaws are wide open. The developed film will show the star pattern with dark centre region (Fig. 6.7). Using a film marker lines are drawn through the middle of the slit images, which will show clearly the intersection point. The lines should intersect with in a ± 2 mm diameter circle. Gantry A ready pack film is sandwiched between two plastic sheets and it is kept on the couch vertically. This means that the plane of the film should be perpendicular to the plane of the couch top. A slit beam is created by moving the jaws optimally, parallel to the gantry axis. The film is exposed for different gantry angles (12 to 30 degree) and the final star pattern is obtained 156 (Fig. 6.8). The lines should intersect with in a ± 2 mm diameter circle.

Quality Assurance Table The above procedure is repeated. The gantry and the collimator is in fixed position. The table is rotated (4-8 times) to different angles and each time the film is exposed. A final star pattern is obtained and it is examined (Fig. 6.9). The lines should intersect with in a ± 2 mm diameter circle.

Fig. 6.7: Mechanical isocenter verification: Collimator rotation Isocenter shift by gantry rotation

The shift is found to be with in 2 mm

Fig. 6.8: Mechanical isocenter verification: Gantry rotation

Isocenter shift by table rotation

The shift is found to be with in 2 mm

Fig. 6.9: Mechanical isocenter verification: Couch rotation

Multiple Beam Alignment Check When more than one beam is used misalignment may occur. This may be due to (i) focal spot displacement, (ii) asymmetry of collimator jaws, and (iii) displacement in the collimator rotation axis or the gantry rotation axis (Lutz et al,1981). To check the beam misalignment a split field test is recommended. A ready pack film is sandwiched between buildup sheet and is exposed twice. First the one half (region 1) of the field is exposed, by blocking the other half (region 2). The gantry is rotated through 180 degree 157

Textbook of Radiological Safety and exposed again by blocking region 1. The relative shift of the two images is the indicator of the misalignment. Photon Beam Data Energy Photon beam energy is specified by the depth dose distribution. A central axis depth dose curve measured with a suitable ion chamber in a water phantom can be compared with published data (BJR 25). The ion chamber should have a small internal diameter (< 3 mm), to minimize displacement correction. It is advisable to compare depth dose ratios for depths beyond dose maximum, instead of absolute values of depth dose. The recommended depth for depth ratios are 10 and 20 cm. The acceptance criteria is specified in terms of depth dose variance for 10 × 10 cm field size,100 SSD, at 10 cm. The acceptable difference is ± 2 % from the published data. The measured depth dose data is to be used for clinical dose calculations. Field Flatness Field flatness for photon beams is defined as the variation of dose relative to the central axis over the central 80 % of the field size at a depth of 10 cm in a plane perpendicular to the central axis. The AAPM -TG 45 specified flatness in terms of maximum percentage variation from the average dose across the central 80 % of the width at half maximum (FWHM) of the profile in a plane transverse to the beam axis. It is given by the relation: M−m × 100 M+m where M and m are the maximum and minimum dose values in the central 80% of the profile. The tolerance limit is ± 3 %. The flatness should be checked for 10 cm and Dmax depths, for maximum field sizes. Beam profiles are generated for inplane, cross plane and diagonal directions and checked for flatness for each given field size. F=

Field Symmetry The profile generated with the above procedure can be used for checking the field symmetry. Usually the profile is folded at the centre and hence the two peripheral halves should be compared at the reference depths. This should not differ more than 2 % at any pair of points located symmetrically with respect to the central ray. Electron Beam Data Energy

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The electron energy is specified by practical electron range (RP) and the most probable energy (EP)O as per AAPM-TG 25. The RP is the depth of the

Quality Assurance point where the tangent to the descending linear portion of the curve intersects the extrapolated background. For range determination one can use ion chambers, diodes or film and RP is found from the depth dose curves. The acceptance limit for the probable energy is ± 0.5 MeV of the nominal energy. In addition, film dosimetry can be used to find R100, R90, R80, and R50. Flatness and Symmetry The flatness of the electron beam is specified in a reference plane perpendicular to the central axis, at the depth of the 95% isodose beyond the depth of dose maximum (AAPM-TG 20). The variation of dose relative to the dose at central axis should not exceed ± 5 % over an area confined within lines 2 cm inside the geometric edge of the fields equal to or larger than 10 × 10 cm. Beam symmetry compares a dose profile on one side of the central axis to that of other side. It should not be more than 2 % any pair of points located symmetrically on opposite sides of the central axis. Other Checks In addition to the above, it is desirable to check the wedge angle (± 2 degree), isocenter shift with couch up and down motion (± 2 mm), optical distance indicator (± 2 mm), field size indicator (± 2 mm), gantry and collimator angles (1degree), laser lights alignment with isocenter (± 2 mm) and table top sag with lateral and longitudinal travel under distributed weight (2 mm) etc. The AAPM-TG 40 has recommended the daily, monthly and annual QA to be carried out in a linear accelerator (Table 6.6). QA FOR HDR BRACHYTHERAPY Electrical and Mechanical Tests After the installation of the HDR unit, the electrical and mechanical tests should be carried out. In the electrical tests, the interlocks in the treatment room doors, guide tube and treatment unit, applicator and guide tubes, emergency stop button to interrupt the irradiation are to be tested. Source safe display and treatment ON/OFF indicator are also be tested. Control console display and control console functions should also be tested for electrical safety. The mechanical tests included are functioning of sensors like pressure, torque, optical and coupling between (i) guide tube and unit, (ii) guide tube and applicator and (iii) drive cable and source etc. Time taken to drive the source to ON/OFF position and integrity of applicators are also very important.

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Textbook of Radiological Safety Table 6.6: QA for Linear accelerators Daily

Monthly

Annual

X-ray out put

X-ray PDD

Electron out put

Electron PDD

Localizing lasers

X-ray beam flatness and symmetry, Electron beam flatness and symmetry, Safety interlocks (emergency switch, wedge and electron cone), Light and radiation field coincidence, Gantry and collimator angle indicators, Wedge and tray position,

Field size dependence of X-ray output, Out put factor constancy for electron applicators, Off axis factor constancy,

Distance indicator Door interlock

Audiovisual monitor

Monitor chamber linearity, X-ray out put vs gantry angle,

Electron out put vs gantry angle, Off axis factor constancy vs gantry angle, Collimator rotation isocenter, Applicator position, Gantry rotation isocenter, Field size indicators, Couch rotation isocenter, Cross hair centering, Coincidence of collimator, gantry, couch axis with isocenter, Couch position indicators, Table top sag, Latching of wedges and tray Vertical travel of table, Jaw symmetry, Safety interlocks, Field light intensity. Arc mode,

Positional Accuracy Applicator Integrity The positional accuracy is tested by combining auto radiography and radiography. A graph sheet is pasted on the therapy verification film. Over the graph sheet the tandem of the standard gynaec applicator (IU3) is fixed. A small lead wire is kept at the end of the tandem in a transverse direction (Fig. 6.10A). The film is exposed with suitable factors with a X-ray machine. This will confirm the mechanical integrity of the applicator. Now the applicator is moved laterally and fixed in a different position in the film. Using the graph sheet the tip of the applicator must be maintained in the same level as before. Then the applicator is connected to the HDR unit in channel 3. Programs are made to create dwell positions at 1500, 1450, 1400 and 1350 (50 mm gap). Autoradiography was performed in HDR for a period of 0.3 s or less at each dwell position. Later the whole test package is taken to X-ray machine and again exposed. When the film exposed, the 160 remaining part of the film was shielded with lead partition. This is repeated for various other applicators.

Quality Assurance Pin Prick Method A graph sheet is pasted on a therapy verification film. The gynec tandem or a flexible implant tube pasted on the graph sheet. A pin is used to mark several cardinal treatment lengths, 1500, 1480, 1460, 1440, 1420 and 1400 mm parallel to the applicator. The distance between the applicator and the pin prick is 50 mm on both sides. The applicator is connected to the transfer tube and the HDR source is used to expose the film. The film was developed and checked for the positioning accuracy and Uniformity. The criteria for positional accuracy is ± 1 mm. Staggered Autoradiography Staggered autoradiography is used to confirm the correct delivery of each unique sequence of dwell positions to the programmed channel, the correct

Fig. 6.10A: Radiography and autoradiography (For color version see plate 1)

Fig. 6.10B: Staggered autoradiography (For color version see plate 2)

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Textbook of Radiological Safety spacing of adjacent dwell positions, and the accurate relative positioning of the most dwell position. Autoradiography is performed for 1-6 channels, with dwell spacing of 15 mm, consisting of 8 dwell positions. The dwell time for each dwell is selected optimally to create a optical density of around one. The first position in each catheter was offset from the previous one by 2.5 mm, resulting in a pattern of descending dots in which errors are readily discernible. In the same way, all the 18 channels has to be tested (Fig. 6.10B). Temporal Accuracy Verification of temporal accuracy consists of identically measuring the length of time the HDR source remains at the specific dwell position and comparing it with the set time. In this study the timer error, timer linearity and end error were determined. Timer Error The HDR system is programed for 60 seconds in 1385 position 21 and the charge for 60 Seconds is noted as R1. Again the system is programed for 60 seconds and the machine is made ON. At 30 Seconds an interruption is made and the system is restarted again. The accumulated charge taking the additional transit time into consideration is noted as R2.Five sets of reading are taken as follows and the average R1 and R2 were found. Reading Charge 1 nC

Charge 2 Charge 3 Charge 4 nC nC nC

Charge 5 nC

Change Average,nC

R1 R2 Timer Error = (R2-R1) t /(2R1-R2)

Timer Linearity The HDR system is programmed for about 320 Seconds at distance 1385 position 21 and the charge for 300 Seconds (T) is noted using an independent timer. Five sets of readings are taken and the average of five sets of reading (Qave) is found and the corrected current ( Icor = Qave/T) is calculated Reading Charge 1 nC

Charge 2 Charge 3 Charge 4 nC nC nC

Charge 5 nC

Change (Qave) nC

R1

Now the system is programmed for 60 sec (Tset) and the charge for 60 seconds is noted by operating the machine timer. Five sets of reading are taken and the average is found. Then the value of the Tmeas is calculated using the formula T =Qave/Icor. Similarly, the readings were taken for 120 162 Sec, 180 Sec, 240 Sec,meas and 300 Sec and the readings are tabulated as follows:

Quality Assurance Tset, Charge1, Charge 2 Charge 3 Charge 4, Charge 5, Average sec nC nC nC nC nC charge (Qave)nC

Tmeas = Qave/Icor

60 120 180 240 300

A graph has been plotted with Tset in X-axis and Tmeas in Y-axis. From the graph the slope is found using the most deviated readings on both sides. Linearity Error = 1 – [Tmeas max (T1) – Tmeas min (T2)] × 100 % Tset corresponds to T1-Tset corresponding to T2 End error is the intercept at Y-axis. Alternatively, the intercept can also be calculated by using Excel Sheet. Leakage and Contamination The HDR machine is connected with a flexible implant tube and made ready for dummy treatment. Trial treatment is performed with the Ir-192 Source for about 5 times. The flexible implant tube is now cut longitudinally and the wipe test is performed using gauze. The gauze is tested for contamination using a contamination Monitor and the reading is recorded as ———CPM. Source Strength Verification The HDR system is to be calibrated with re-entrant type ion chamber (well chamber). As a first step the axial response of the chamber is obtained, by programing all dwell positions (say1-48). The secondary standard dosimeter is used to measure the current in nA. Repeated readings are taken for at least 2 times for a given dwell position and tabulated as shown below. A plot is made between dwell position and relative meter readings. The dwell position (Dwellmax) corresponding to the maximum response is found. Dwell position

Meter reading, nA (i)

Meter reading, nA (ii)

Now the machine set for the dwell position of Dwellmax and the measurements are repeated for 3 times. The readings are tabulated as given 163

Textbook of Radiological Safety below and the average of the readings is also found. The temperature (T) and pressure (P) parameters are also noted. Dwell position Meter reading, Meter reading, Meter reading, Average Meter nA (i) nA (ii) nA (iii) reading ,nA (M) Dwellmax

Activity = M × N × Ktp where Ktp is the temperature and pressure correction factor = ((273.15 + T)/(273.15 + 22)) * 1013/P) and N is the Chamber calibration factor = (

) GBq/nA

The % of variation of the measured activity with that of the stated activity (ventors) is calculated as follows: % variation =

Stated activity Measured activity ×100 Stated activity

QA for Treatment Planning System Quality assurance of treatment planning system (TPS) is an essential and indispensable part before commissioning any TPS. After installation the TPS is tested for its accuracy of (i) digitization of coordinates, (ii) data transfer from orthogonal radiographs, (iii) dose calculation at selected anatomical points, and (iv) computational algorithm etc. A 5 cm, 10 cm and 12 cm catheter length is drawn in a graph sheet and its AP and lateral images are fed into the system through the digitizer. A printout is obtained and its dimensions are compared with input data. Similarly input is made through keyboard and the hard copy is compared. This can be repeated for different catheter lengths and orientation with marking points. The tolerance limit for data transfer is ± 1 mm. A single pellet is programmed in a gynaec application, so that it can deliver a dose of 10 Gy at 10 mm distance. The isodose print out is made and checked. The treatment time can be verified by either by (i) Mante carlo data or (ii) Meisberger polynomial. The tolerance limits is ± 3 % for dose. Periodic QA schedule Four flexible catheters are taped on a sheet and placed on the ready pack film at 2 cm apart. Dummies were inserted in each catheter, making sure that the stop collar on the dummies abuts the coupler on each catheter (Fig. 6.11). After radiographing the dummies, the HDR machine is 164 programmed to dwell 0.1 sec at each dwell position. At each catheter, dwell

Quality Assurance

Fig. 6.11: Periodic quality assurance test (For color version see plate 2)

positions are 1, 5, 9, 13, 17, 21, 25, 29, 33, 37, 41, 45, 48. After the exposure, the HDR machine is again programmed as follows: Catheter 1 Dwell 5 Length 1490 Catheter 2 Dwell 5 Length 1450 Catheter 3 Dwell 5 Length 1400 Catheter 4 Dwell 5 Length 1390 After HDR autoradiography, the film is developed and the maximum deviation from the dummy position is checked and the tolerance is 2 mm. BIBLIOGRAPHY 1. Advances in Diagnostic Medical physics Proceedings of the international symposium on advances in diagnostic Medical physics and workshop on Cyclotron PET/CT, July 13-16, 2006, Edited by GS Pant, Himalaya publishing house, Delhi. 2. Comprehence QA for radiation oncology: Report of AAPM radiation therapy Task Group 40. Med Phys1994;21:581-618. 3. Faiz M.Khan: The Physics of radiation therapy, (3rd edn.) Lippincott Williams & Wilkins, 2003. 4. Jeffrey W, Zuoferg Li; Mate carlo dosimetry of the Microselectron pulsed and high dose rate Ir-192 sources. Med phys 1995;22(6):809-19.

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Textbook of Radiological Safety 5. Nath R, Biggs PJ, Bova FJ, et al. AAPM code of practice for radiotherapy accelerators: report of AAPM radiation therapy Task Group No. 45. Med Phys 1994;21:1093-1121. 6. Nuclear Medicine resources book, IAEA, Vienna 2006. 7. Performance evaluation of the new whole-body PET/CT scanner: Discovery ST: Valentino Bettinardi et.al. European Journal of Nuclear Medicine and Molecular Imaging Vol. 31, No. 6, 2004;867-81. 8. QA Instructions to users: Nucletron India, No 3,D’silva road, Mylapore,Chennai600 004. 9. Thayalan K. Physical and dosimetric studies of High dose rate Brachytherapy system with clinical correlation, in carcinoma of the uterine cexvix-PhD thesis. The Tamilnadu Dr MGR Medical university, Chennai 2003.

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7

Regulations and Dose Limits

ATOMIC ENERGY ACT-1962 The primary legislation to regulate the use of ionizing radiation, in India is the Atomic Energy Act,1962. The Act and the secondary legislation, viz. Radiation Protection Rules 1971 (RPR-1971) provides necessary regulatory infrastructure for effective implementation of radiation protection program. The RPR-1971 was revised in 2004 and named as Atomic Energy (Radiation Protection) Rules, 2004. The Act empowers the Government of India to exercise control over protection and the use of Atomic energy. It offers special provisions to safety, under section 17 and powers to make rules under section 30 of the Act. The Act envisage control over premises where radioactive substances are handled or radiation generating equipments are operated. The Act emphasizes safety while working with radiation. It deals with control over possession, use, sale and transport of radioactive materials and cognizance of offences. The legislation and various rules related to radiation protection in medicine are listed below. Most of them are available in the form of codes and guides: 1. Atomic Energy Act 1962 (33 of 1962) 2. Atomic Energy Regulatory Board, 1983 3. Atomic energy (Radiation protection) Rule 2004 4. Safe disposal of Radioactive Waste Rules, 1987 5. Radiation Surveillance procedures for medical applications-1989 6. AERB Safety code SC/MED-1, Telegamma therapy equipment and installation, 1986 7. AERB Safety code SC/MED-2, Medical diagnostic X-ray equipment and installations, 2001 8. AERB Safety code SC/MED-3, Brachytherapy sources, equipment and installation, 1988 9. AERB Safety code SC/MED-4(rev.1), Nuclear medicine facilities, 2001 10. AERB Safety code SC/TR-1, Transport of radioactive material. ATOMIC ENERGY REGULATORY BOARD The Atomic Energy Regulatory Board (AERB), was constituted on November 15, 1983, by the President of India by exercising the powers conferred by section 27 of the Atomic Energy Act 1962 (33 of 1962). It is an

Textbook of Radiological Safety apex body that regulates the use of ionizing radiation in the country. The AERB is entrusted with the responsibility of developing and implementing appropriate regulatory measures aimed at ensuring radiation safety in applications involving ionizing radiations. The board is fully empowered to lay down standards and frame rules and regulations. The chairman AERB is the competent authority, recognized by the Government for enforcing provisions of radiation safety in the use of ionizing radiation. AERB has jurisdiction over all the units of the department of Atomic energy and all radiation installations in the country. The mission of the board is to ensure that the use of ionizing radiation and nuclear energy does not cause undue risk to health and environment. The board covers the safety aspects of all areas of nuclear fuel cycle and use of radiation in medicine, agriculture, industry and research and transport of radioactive materials. The board is assisted by Health, safety and environment group of BARC viz. Radiological physics and advisory division (RPAD), advisory committees and task groups. The major objectives of AERB is to develop and publicize specific codes and guides, which will deal with the radiation safety aspects of various applications of radiations. It will also issue authorization related to site, design, manufacture, construction, commissioning, operation, maintenance, and decommissioning and disposal of radioactive sources. RADIATION PROTECTION RULES-2004 [Published in the Gazette of India: September 11, 2004] Part-II-Section 3-Sub-section (I) Government of India Department of Atomic Energy Mumbai, the 25th August, 2004 G.S.R. 303.— In exercise of the powers conferred by Section 30 read with section 3 and clause (i) and sub-clauses (c) and (d) of clause (ii) of SubSection (1), Sub-section (4) of Section 14, and Sections 16, 17 and other relevant Sections of the Atomic Energy Act (33 of 1962) and all other powers enabling it in this behalf, and in supercession of Radiation Protection Rules 1971 except as respects things done or omitted to be done before such supercession, the Central Government hereby makes the following rules, namely:Rule 1. Short Title, Extent and Commencement 1. These rules may be called the Atomic Energy (Radiation Protection) Rules, 2004. 2. These rules shall apply to practices adopted and interventions applied with respect to radiation sources. 168

Regulations and Dose Limits 3. They extend to the whole of India. 4. They shall come into force from the date of their final publication in the Official Gazette. Rule 2. Definitions Define the various terms and terminology used in the Atomic energy Act and ARPR 2004. Rule 3. Licence 1. No person shall, without a licence (a) establish a radiation installation for siting, design, construction, commissioning and operation; and (b) decommission a radiation installation. 2. No person shall handle any radioactive material, or operate any radiation generating equipment except in accordance with the terms and conditions of a license. (Radiation installation in medicine includes the (i) the medical X-ray equipments, (ii) radiation therapy equipments, and (iii) nuclear medicine equipments. Any institution desire to start a radiation facility has to obtain License, authorization, registration, consent from AERB). 3. A license shall be issued for sources and practices associated with the operation of i. Nuclear fuel cycle facilities; ii. Land based high intensity gamma irradiators other than gamma; irradiation chambers; iii. Particle accelerators used for research and industrial applications; iv. Neutron generators; v. Facilities engaged in the commercial production of radioactive material or radiation generating equipment; vi. Telegamma and accelerators used in radiotherapy; vii. Computed tomography (CT) unit; viii. Interventional radiological X-ray unit; ix. Industrial radiography; and x. Such other source or practice as may be notified by the competent authority, from time to time. Provided that for sources and practices associated with the operation ofi. Brachytherapy; ii. Deep X-ray units, superficial and contact therapy X-ray units; iii. Gamma irradiation chambers; iv. Nuclear medicine facilities; v. Facilities engaged in the commercial production of nucleonic gauges and consumer products containing radioactive material; and vi. Such other source or practice as may be notified by the competent authority, from time to time; an authorization shall be necessary. 169

Textbook of Radiological Safety i. Provided further that for sources and practices associated with the operation of ii. Medical diagnostic X-ray equipment including therapy simulator; iii. Analytical x-ray equipment used for research; iv. Nucleonic gauges; v. RIA laboratories; vi. Radioactive sources in tracer studies; vii. Biomedical research using radioactive material; and viii. Such other source or practice as may be notified by the competent authority, from time to time; a registration shall be necessary. Provided also that for i. Approval for siting, design, construction, commissioning and decommissioning of a radiation installation; ii. Approval for sealed sources, radiation generating equipment and equipment containing radioactive sources, for the purposes of manufacture and supply; iv. Approval for package design for transport of radioactive material; v. Approval for shipment approval for radioactive consignments; and vi. Such other source or practice as may be notified by the competent authority, from time to time; consent shall be necessary. 4. The license shall not be transferable without the prior approval of the competent authority. Rule 4. Fees for License The competent authority may prescribe by notification in the Official Gazette, appropriate fees payable for issuance of license specified in these rules. Rule 5. Exemption The use and disposal of an substance and materials which spontaneously emit radiation not exceeding the level of radiation prescribed by notification issued under clause (i) of Sub-section (1) of Section 2 of the Act and the use of radiation generating equipment, devices or appliances emitting radiation not exceeding the limit determined by the Central Government under clause (g) of Section 3 of the Act, are exempted from the purview of rule 3. Rule 6. Exclusion Exposures resulting from naturally occurring radionuclides present in the human body, cosmic radiation at the earth surface, unmodified 170 concentrations of radionuclides in raw materials and from other sources

Regulations and Dose Limits and practices which may be prescribed as not amenable for control, are excluded from these rules. Rule 7. Conditions Precedent to the Issuance of a License 1. An application for license shall be made to the competent authority by an employer or a person duly authorized by him. 2. No license to handle radioactive material, or to operate radiation generating equipment, shall be issued to a person unless, in the opinion of the competent authority a. The application for such license is for purposes envisaged by the Act; b. Documentation relevant to the license and complete in all respects is submitted to the competent authority; c. In respect of approval for siting, design, construction, commissioning and decommissioning, of a radiation installation, the proposed equipment, facilities and handling procedures afford adequate protection during normal or intended operations; d. The applicant has demonstrated compliance with the provisions of the relevant safety codes and safety standards specified by the competent authority; and e. In respect of license for operation of a radiation installation. i. All the requirements relating to safety specified by the competent authority in the relevant safety codes and safety standards have been satisfied in the construction of the radiation installation; ii. Workers have appropriate training and instructions in radiation safety, in addition to the appropriate qualification and training required for performing their intended tasks; iii.A Radiological Safety Officer is designated in accordance with rule 19; iv.Appropriate radiation monitors and dosimetry devices are available with the applicant for purposes of radiation surveillance; v. the equipment, facilities and handling procedures afford adequate protection during normal operations, minimize occurrence of potential exposures and enable appropriate remedial actions to be taken in the event of an accident. 3. No type approval of sealed sources, radiation generating equipment and equipment containing a radioactive source for the purpose of manufacture and supply or package design approval for transport of radioactive material or shipment approval for radioactive consignment or any other approval as notified under third proviso to rule 3, by the competent authority may be issued unless, in the opinion of the competent authority, the applicant has demonstrated compliance with the relevant safety codes and safety standards specified by him.

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Textbook of Radiological Safety Rule 8. Issuance of License The license shall be issued within a period of one hundred and eighty days from the date of receipt of the application subject to the condition that all the requirements for issuance of the license have been duly fulfilled. Rule 9. Period of Validity of License Every license issued under rule 3 shall, unless otherwise specified, be valid for a period of 5 years from the date of issue of such license. Rule 10. Suspension, Modification or Withdrawal of a License The competent authority may i. If in its opinion, the licensee has contravened any of the provisions of these rules; or ii. Considers it to be necessary in public interest pertaining to radiation safety; after giving a show cause notice to the licensee and also giving him an opportunity to make a representation within a period of thirty days from the date of receipt of the notice by him against the action proposed to be taken and on consideration of his representation, a. Suspend the operation of the license for a specified period of time; or b. Revoke or modify the terms and conditions of the license. Rule 11. Modification of Radiation Installation or Change in Working Condition No modification to an existing radiation installation or no change in working conditions therein, affecting safety shall be done without the prior approval of the competent authority. Rule 12. Restrictions on Use of Sources 1. The licensee shall not handle any source:a. Other than those specified in the license; b. For any purpose other than those specified in the license; and c. In any location except as specified in the license. 2. The licensee shall ensure that individuals other than those who may be specified in the license do not handle the source. Rule 13. Restriction on Certain Practices 1. Practices such as deliberate addition of radioactive substances in foodstuffs, beverages, toys, personal ornaments, and cosmetics or any other commodity or product intended for ingestion, inhalation or percutaneous intake by, or application to, a human being and sale, import or export of such products shall not be permitted. 172 2. Activation of the aforesaid products shall not be permitted.

Regulations and Dose Limits Rule 14. Radiation Symbol or Warning Sign 1. The radiation symbol or warning sign shall be conspicuously and prominently displayed at all times a. On externally visible surfaces of radiation equipment, and containers for storage of radioactive materials; packages for radioactive materials and vehicles carrying such packages; b. At the entrance to the room housing the radiation generating equipment; and c. At the entrance of controlled area and supervised area. 2. The radiation symbol shall not be used for any purpose other than those mentioned in these rules. 3. The specification of the radiation symbol or warning sign shall be as prescribed by the competent authority, by order for that purpose. Rule 15. Dose Limits and Other Regulatory Constraints The licensee shall ensure compliance with the dose limits and other regulatory constraints specified by the competent authority by order under these rules. Rule 16. Safety Standards and Safety Codes The competent authority may issue safety codes and safety standards, from time to time, prescribing the requirements for radiation installation, sealed sources, radiation generating equipment and equipment containing radioactive sources, and transport of radioactive material and the licensee shall ensure compliance with the same. Rule 17. Prohibition of Employment of Persons below Certain Age 1. No person under the age of 18 years shall be employed as a worker. 2. No person under the age of 16 years shall be taken as trainee or employed as an apprentice for radiation work. Rule 18. Classified Worker The employer shall designate as classified workers, those of his employees, who are likely to receive an effective dose in excess of three tenths of the average annual dose limits notified by the competent authority and shall forthwith inform those employees that they have been so designated. Rule 19. Radiological Safety Officer Every employer shall designate, with the written approval of the competent authority, a person having appropriate qualifications as Radiological Safety 173 Officer.

Textbook of Radiological Safety Rule 20. Responsibilities of the Employer 1. Every employer shall: a. Ensure that provisions of these rules are implemented by the licensee, Radiological Safety Officer and other worker(s), b. Provide facilities and equipment to the licensee, Radiological Safety Officer and other worker(s) to carry out their functions effectively in conformity with the regulatory constraints, c. Prior to employment of a worker, procure from his former employer, where applicable, the dose records and health surveillance reports, d. Upon termination of service of worker provide to his new employer on request his dose records and health surveillance reports, e. Furnish to each worker dose records and health surveillance reports of the worker in his employment annually, as and when requested by the worker and at the termination of his service, f. Inform the competent authority if the licensee or the Radiological Safety Officer or any worker leaves the employment, and g. Arrange for health surveillance of workers as specified under rule 25. 2. The employer shall be the custodian of radiation sources in his possession and shall ensure physical security of the sources at all times. 3. The employer shall inform the competent authority, within 24 hours, of any accident involving a source or loss of source of which he is the custodian. Rule 21. Responsibilities of the Licensee 1. The responsibility for implementing the terms and conditions of the license shall rest with the licensee. 2. The licensee shall comply with the surveillance procedures, safety codes and safety standards specified by the competent authority. 3. Every licensee shall establish written procedures and plans for controlling, monitoring and assessment of exposure for ensuring adequate protection of workers, members of the public and the environment and patients, wherever applicable. 4. The licensee shall comply with the provision of rules for safe disposal of radioactive waste issued under the Act. 5. Without prejudice to the generality of the above, the licensee shall a. Not allow workers, other than those specified in sub-clause (ii) of clause (e) of sub-rule (2) of rule 7 and already dealt with under rule 17. b. Maintain records of workers as specified under rule 24; c. Arrange for preventive and remedial maintenance of radiation protection equipment, and monitoring instruments; d. In consultation with the Radiological Safety Officer, investigate any case of exposure in excess of regulatory constraints received by individual workers and maintain records of such investigations; 174

Regulations and Dose Limits e. Inform competent authority promptly of the occurrence, investigation and follow-up actions in cases of exposure in excess of regulatory constraints, including steps to prevent recurrence of such incidents; f. Carry out physical verification of radioactive material periodically and maintain inventory; g. Inform appropriate law enforcement agency in the locality of any loss of source; h. Inform the employer and the competent authority of any loss of source; i. Investigate and inform the competent authority of any accident involving source and maintain record of investigations; j. Verify the performance of radiation monitoring systems, safety interlocks, protective devices and any other safety systems in the radiation installation; k. In consultation with Radiological Safety Officer, prepare emergency plans, as specified in rule 33, for responding to accident to mitigate their consequences and ensure emergency preparedness measures; l. Conduct or arrange for quality assurance tests of structures, systems, components and sources and related equipment; m.Advise the employer about the modifications in working condition of a pregnant worker; n. Inform the competent authority if the Radiological Safety Officer or a worker leaves the employment; and o. Inform the competent authority when he leaves the employment. 6. The licensee shall ensure that the workers are familiarised with contents of the relevant surveillance procedures, safety standards, safety codes, safety aides and safety manuals issued by the competent authority and emergency response plans. Rule 22. Responsibilities of the Radiological Safety Officer 1. The Radiological Safety Officer shall be responsible for advising and assisting the employer and licensee on safety aspects aimed at ensuring that the provisions of these rules are complied with. 2. The Radiological Safety Officer shall:a. Carry out routine measurements and analysis on radiation and radioactivity levels in the controlled area, supervised area of the radiation installation and maintain records of the results thereof; b. Investigate any situation that could lead to potential exposures; c. Advise the employer regarding i. The necessary steps aimed at ensuring that the regulatory constraints and the terms and conditions of the license are adhered to; ii. The safe storage and movement of radioactive material within the radiation installation;

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Textbook of Radiological Safety iii. Initiation of suitable remedial measures in respect of any situation that could lead to potential exposures; and iv. Routine measurements and analysis on radiation and radioactivity levels in the off-site environment of the radiation installation and maintenance of the results thereof. d. Ensure that i. Reports on all hazardous situations along with details of any immediate remedial actions taken are made available to the employer and licensee for reporting to the competent authority and a copy endorsed to the competent authority; ii. Quality assurance tests of structures, systems, components and sources, as applicable are conducted; and iii. Monitoring instruments are calibrated periodically. e. Assist the employer in i. Instructing the workers on hazards of radiation and on suitable safety measures and work practices aimed at optimizing exposures to radiation sources; and ii. The safe disposal of radioactive wastes; and iii. Developing suitable emergency response plans to deal with accidents and maintaining emergency preparedness. f. Advise the licensee on i. The modifications in working condition of a pregnant worker; and ii. The safety and security of radioactive sources. g. Furnish to the licensee and the competent authority the periodic reports on safety status of the radiation installation; and h. Inform the competent authority when he leaves the employment. Rule 23. Responsibilities of Worker 1. Every worker shall observe the safety requirements and follow safety procedures and instructions and shall refrain from any wilful act that could be detrimental to self, co-workers, the radiation installation and public. 2. The worker shall:a. Provide to the employer information about his previous occupations including radiation work, if any; b. Make proper use of such protective equipment, radiation monitors and Personnel monitoring devices as provided; and c. Inform the licensee and the Radiological Safety Officer, of any accident or potentially hazardous situation that may come to his notice; 3. A female worker shall, on becoming aware that she is pregnant, notify the employer, licensee and Radiological Safety Officer in order that her working conditions may be modified, if necessary.

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Regulations and Dose Limits Rule 24. Records of Workers 1. Every licensee shall maintain complete and up-to-date records of a. personnel monitoring under Clause (b) of sub-rule (2) of rule 27, in the format as specified by order by the competent authority; and b. the health surveillance specified in rule 25. 2. Such records shall be preserved during the working life of each worker, and afterwards until the worker attains or would have attained the age of 75 years, or not less than 30 years after the termination of the work involving occupational exposure whichever is later. 3. A worker shall have access to his personnel monitoring and the health surveillance records. Rule 25. Health Surveillance of Workers 1. Every employer shall provide the services of a physician with appropriate qualifications to undertake occupational health surveillance of classified workers. 2. Every worker, initially on employment, and classified worker, thereafter at least once in three years as long as the individual is employed, shall be subjected to the following a. general medical examination as specified by order by the competent authority; and b. health surveillance to decide on the fitness of each worker for the intended task; 3. The health surveillance shall include a. special tests or medical examinations as specified by order by the competent authority, for workers who have received dose in excess of regulatory constraints; and b. counseling of pregnant workers. Rule 26. Medical Exposures The licensee carrying out diagnostic or therapeutic work using radiation generating equipment, sealed or unsealed sources, shall for optimizing the medical exposure ensure that a. Performance of the equipment is verified periodically by appropriate quality assurance tests; b. Records are maintained for a period specified by the competent authority of i. radiation doses received by therapy patients; ii. activity administered to patients for diagnostic and therapeutic purposes; and iii.other relevant parameters. c. The exposure of humans for biomedical research is carried out only on healthy volunteers with their prior consent in writing. The methodology, 177

Textbook of Radiological Safety the number of volunteers and the radiation dose they are subjected to shall be reviewed by the ethical review committee constituted by the employer; and d. Any accidental medical exposure is investigated and a written report is submitted to the competent authority. Rule 27. Radiation Surveillance Requirements 1. The competent authority may by order specify appropriate radiation surveillance requirements and procedures and the employer and the licensee shall comply with them. 2. Without prejudice to the generality of the foregoing provisions, such radiation surveillance requirements and procedures may provide that a. the siting, design, construction, commissioning, operation, servicing and maintenance and decommissioning of facilities involving the use of radiation, and disposal of radioactive material shall be done in accordance with the specifications laid down by the competent authority in the relevant safety codes and safety standards; b. the workers shall be subjected to personnel monitoring and health surveillance and appropriate records shall be maintained; c. transport of radioactive material in public domain shall be in accordance with the procedures laid down by the competent authority and in accordance with the other regulations pertaining to transport by different modes; and d. appropriate quality assurance requirements in the above. Rule 28. Directives in the Cases of Exposures in Excess of Regulatory Constraints a. When, in the opinion of the competent authority, any worker has exceeded the dose constraints, the competent authority may, without prejudice to other course of action available, issue appropriate directives for controlling further exposure and the employer shall comply with the directives. b. If a worker discontinues radiation work under the directives of the competent authority issued under this rule, the employer shall assign alternative work not involving exposure to radiation, until the competent authority is satisfied about the fitness of the worker to resume radiation work. c. The employer shall comply with restrictions, if any, that the competent authority may impose in this regard. Rule 29. Power to Appoint or Recognize Persons or Agencies The competent authority may, from time to time, appoint or recognize

178 persons or agencies having the qualifications and expertise, prescribed in

Regulations and Dose Limits the relevant safety code, for the purpose of performing any of the functions entrusted to them by the authority and for ensuring compliance with radiological surveillance. Rule 30. Inspection of Premises, Radiation Installations and Conveyances 1. Any person duly authorized under Sub-section (4) of Section 17 of the Act may, for the purposes of enforcement of these rules, inspect any premises, or radiation installation, or conveyance. 2. The date and time of inspection may or may not be informed to the employer or the licensee prior to the inspection. 3. The employer and the licensee shall extend all assistance to enable the inspection to be carried out effectively and unhindered. 4. The findings of the inspection shall be forwarded to the licensee for necessary corrective actions. 5. Inspection may be carried out at all licensing stages, namely, siting, construction, commissioning, operation and decommissioning. 6. The person authorised to conduct inspection may – a. Inspect, from safety point of view, to ensure that the licensee has fulfilled the radiological safety requirements for carrying out the practices at the radiation installation as per the stipulations laid down in the licence. This shall include i. Checking, whether the safety related structures, systems, components and devices are of approved quality based, on the relevant safety codes and safety standards specified by the competent authority and that they are functioning as per the design intent, checking that respective operating personnel are competent to operate the facility; ii. That the facilities are operating as per the approved technical specification; and iii. Conducting all such examinations (including verification of relevant records) as may be considered necessary. b. Make such tests and measurements as may be necessary for the purpose of assessing radiation safety; c. Investigate unusual incidents or accidents, if any, that had occurred at the radiation installation and arrive at the reasons for the same and recommend corrective measures; d. Review and verify whether the corrective actions have been implemented; and e. Inspect radioactive consignments in any conveyance carrying radioactive material and inspect any package containing radioactive material.

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Textbook of Radiological Safety Rule 31. Power to Investigate, Seal or Seize Radiation Installation or Radioactive Material and to give Direction to the Employer 1. Any person duly authorised under Section 17 of the Act, may, after inspection, carry out investigation for the purposes of determining contravention of any of the provisions of these rules; 2. The investigation may be carried out against a complaint or on suspicion or after an unusual incident or accident; 3. The person authorised to investigate may a. Seal any radiation installation or any conveyance carrying radioactive materials or seize any radioactive material or contaminated equipment; and b. Indicate in writing to the employer any recommendation aimed at ensuring adequate protection and the licensee shall comply with the same. Rule 32. Directives in Case of Accidents 1. In the event of an accident involving the source or release of radioactive material, the competent authority may a. Intervene and issue such directions as deemed fit and proper under the circumstances in the interest of radiation safety and the employer shall act as per the directions of the competent authority and shall make every effort to mitigate the consequences of the accident, or b. The competent authority may assign experts to give advice or render assistance in mitigating the consequences of the accident and the expenses incurred, if any, shall be reimbursed by the employer. 2. In the interest of safety of the radiation installation, workers, public and the environment, the competent authority may issue such directions as it may deem fit for ensuring safety including the immediate shutting down of the radiation installation and the employer shall comply with the directions. Rule 33. Emergency Preparedness 1. The licensee shall prepare emergency response plans as specified by the competent authority in the relevant safety codes and maintain emergency preparedness. 2. The licensee shall submit the response plans for plant emergencies and site emergencies to the competent authority for approval. 3. The licensee shall submit the response plans for off-site emergencies prepared by the appropriate authorities to the competent authority for review. 4. In respect of radiation installations governed by clause (a) of sub-rule 180 (3) of rule 3 and clause (b) of sub-rule (3) of rule 3, emergency response

Regulations and Dose Limits plans shall be submitted to the competent authority prior to the commissioning of the installations. 5. Any modification to the emergency plan shall require prior approval of or review by the competent authority. Rule 34. Decommissioning of Radiation Installation 1. When a radiation installation or radiation generating equipment ceases to be in use, the employer shall ensure its decommissioning. 2. No employer shall decommission a radiation installation without the prior approval of the competent authority. 3. The decommissioning plan shall take due cognizance of disposal of radioactive wastes, recycling of materials, and reuse of equipment and premises. 4. The licensee shall comply with such directive as may be issued by the competent authority to ensure adequate protection of the persons in and around the decommissioned installation. Rule 35. Offences and Penalties Any person who contravenes the provisions of these rules or any of the terms and conditions of license issued hereunder, shall be punishable as provided for under the Act. – [F.No. AEA/30(1)/2002-ER],V.P. RAJA. Jt. Secy. REGULATORY CONTROLS FOR DIAGNOSTIC X-RAY EQUIPMENT AND INSTALLATIONS 1. Design certification: Every medical diagnostic X-ray equipment shall meet the design safety specifications stipulated in the safety code (SC/ MED/-2(Rev.1)).The manufacturer/vendor shall obtain design certification from the competent authority prior to manufacturing the X-ray equipment. 2. Type approval /No objection certificate: Prior to marketing the X-ray equipment the manufacturer shall obtain a Type approval certificate from the competent authority for indigenously made equipment. For equipment of foreign make, the importing /vending agency shall obtain a No Objection Certificate (NOC) from the competent authority, prior to marketing the equipment. Only Type approved and NOC validated equipment shall be marketed in the country. 3. Approval of layout: No X-ray unit shall be commissioned unless the layout of the proposed X-ray installation is approved by the competent authority. The application for approval shall be made by the person owning responsibility for the entire X-ray installation. 4. Registration of X-ray equipment: Acquisition of an X-ray equipment, by purchase, transfer, gift, leasing or loan, shall be registered with the 181 competent authority by the person acquiring the equipment.

Textbook of Radiological Safety 5. Commissioning of X-ray equipment: No X-ray equipment shall be commissioned unless it is registered with the competent authority. 6. Inspection of X-ray installations: The diagnostic X-ray installations shall be made available by the employer/owner for inspection, at all reasonable times, to the competent authority or its representative, to ensure compliance with the safety code. 7. Decommissioning of X-ray installations: Decommissioning of X-ray equipment shall be registered with the competent authority immediately by the employer /owner of the equipment 8. Certification of RSO: Any person accepting assignment to discharge the duties and functions of RSO in diagnostic X-ray installations shall do so only after obtaining certification from the competent authority for the purpose. Such certification shall be granted on the basis of adequacy of the persons qualifications, experience and testing /survey / dosimetry equipment available 9. Certification of service engineers: Only persons holding valid certificate from the competent authority shall undertake servicing of X-ray equipment. Certification shall be granted on the basis of qualifications, training, experience and safety record of such person and availability of servicing facilities. Any person who contravenes the provisions of these rules or any other terms or conditions of license/registration /certification granted to him/her by the competent authority, shall be punishable under sections 24,25 and 26 of the Atomic Energy Act,1962. The punishment may include fine, imprisonment, or both, depending on the severity of the offence. Personnel Requirements Every hospital shall have a qualified RSO (either full time or part time), Radiologist, X-ray technologist and a service engineer. He should be delegated with the responsibility of ensuring radiation safety applicable to the installation. The minimum qualification and experience required are given the safety code AERB/SC/MED-2 (Rev.1). Personnel Responsibilities Manufacturer

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The manufacturer /vendor of the equipment shall make available to the user the procedures to QA tests, exposure charts, operating manuals and a copy of safety documents issued by the competent authority from time to time, and maintenance facilities during useful life time of X-ray equipment. In the case of CT scan, the manufacturer shall provide the required phantoms for dosimetry and image quality checks.

Regulations and Dose Limits Service Engineer The service engineer undertaking services in a radiological installation shall immediately report to the competent authority, about the equipment, which is no longer safe for use. He should furnish brief description of the equipment, its location, address, the name and address of the owner and the nature of defects that make the equipment hazardous. Employer The employer should ensure the availability of qualified RSO and qualified personnel for handling the X-ray equipment. He should also provide the required equipments and facilities to discharge their duties. The employer shall also be responsible for ensuring that personnel monitoring devices are made available to the radiation workers. The employer should ensure that persons handling medical X-ray equipment are duly abide by the provisions of the safety code. He should also ensure the availability of safety codes, issued by the competent authority to the workers. Radiologist The radiologist shall undertake an X-ray examination on the basis of medical requirement. He/she so conduct the examination as to achieve maximum reduction in radiation dose to the patient while retaining all clinically important information. X-Ray Technologist X-ray technologist and other attending staff shall ensure appropriate patient protection, public protection and operational safety in handling X-ray equipment and other associated facilities. Student/Trainee Medical students/trainees shall not operate X-ray equipment except under direct supervision of authorized operating personnel. Radiological Safety Officer (RSO) The RSO assist the employer to fulfill the regulatory requirements, applicable to that installation. He shall implement all radiation surveillance measures, conduct periodic radiation protection surveys, maintain proper records of periodic QA tests and personnel doses, instruct all workers on relevant safety measures, educate and train new entrants, and take local measures. This includes the issuance of administrative instructions in writing, to deal with radiation emergencies. The RSO shall ensure that all radiation measuring and monitoring instruments in his custody are properly calibrated and maintained in good 183

Textbook of Radiological Safety condition. Suitable records of such surveys, including layout drawings, dose mappings, deficiencies noticed and remedial actions taken, shall be maintained for future follow up. REGULATORY CONTROLS FOR NUCLEAR MEDICINE FACILITIES Consent

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1. The consent for any practice involving radiation exposure is based on a system of notification, registration, authorization, license or exemption from regulatory control as established by the competent authority. 2. The consent is issued on the basis of written application. The consentee shall ensure that persons handling radioactive materials for nuclear medicine purposes are familiar with the mandatory provisions of RPR 2004. Radiation surveillance procedures for Medical applications of radiation, 1989. Radiation surveillance procedures for transport of Radioactive material, 1988, Atomic energy (safe disposal of radioactive waste) rules 1987, and safety directives issued by the competent authority from time to time, and other instructions of the competent authority in specific cases. The consentee shall ensure compliance with the mandatory requirements specified in the above documents. 3. The consentee shall designate with the approval of the competent authority a person having suitable qualifications, to function as RSO. Nuclear medicine facilities carrying out diagnostic in-vivo/in-vitro investigations shall an RSO of level-II ,whereas facilities where therapy is also carried out shall have RSO of level-III. 4. Authorization from competent authority is required to procure radioactive material. The consentee shall be solely responsible for the safety and security of the radioactive material, its proper use, and the safe disposal of wastes. The consentee shall maintain an up to date inventory and account for decay and disposal of sources. 5. The nuclear medicine facility shall not be commissioned until the competent authority approves the facility. Any change or modification to an already approved facility shall be carried out only with the prior approval/sanction of the competent authority. 6. Transport of radioactive material in public domain shall be in accordance with the provisions of code AERB/SC/TR-1. 7. The consentee shall not take the radioactive material out of the approved premises. The consentee shall not lend, gift, transfer, or sell any radioactive material and shall not receive radioactive material other than those specified in the authorization. 8. Radioactive material shall not be disposed off without prior approval of the competent authority.

Regulations and Dose Limits 9. Any person duly authorized by the competent authority may inspect the nuclear medicine facility, the radioactive sources available in house, the radioisotope inventory, logbooks, records of area monitoring and contamination monitoring, instruments and devices used for the above and / or quality assurance programme, records of unusual occurrences during handling of the sources and transport of radioactive materials. If, on inspection, any evidence of non compliance of any of the mandatory provision is noticed, the inspectors may a. Seal the institution or transport or conveyance carrying radioactive materials or seize radioactive material or contaminated equipment, and b. Indicate in writing to the consentee any modification aimed at providing adequate protection, and the consentee shall comply with the same 10. The nuclear medicine facility shall be decommissioned only after prior approval of the competent authority and after all radioactive and contaminated materials from the facility. Prior approval of the competent authority is necessary for reuse of the facility. 11. Any person who contravenes the provisions of these rules or any other terms and conditions of approval granted to him/her by the competent authority, shall be punishable under sections 24,25 and 26 of the Atomic Energy Act, 1962. The punishment may include fine, imprisonment, or both, depending on the severity of the offence. Personnel Requirements Every nuclear medicine department shall have a qualified nuclear medicine physician, nuclear medicine technologist, and a RSO either Level-II or LevelIII. The minimum qualification and experience required are given the safety code AERB/SC/MED-4 (Rev.1). Personnel Responsibilities Consentee The consentee shall employ adequate number of personnel, provide appropriate equipment and tools to concerned persons for safe handling of radioactive material. He shall also provide personnel monitoring devices to the radiation workers. He shall constitute a local safety committee to review the operational safety, quality assurance, ethical aspects and regulatory compliance. He should report the competent authority about the safety committee, change in safety and unusual occurrence if any. He should also ensure the availability of safety codes, issued by the competent authority to the workers.

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Textbook of Radiological Safety Nuclear Medicine Physician i.

ii.

iii.

iv.

v.

The efficacy of new nuclear medicine procedures is based on the experience of previous clinical trails, animal experiments and published literature. Prior clearance from the nuclear medicine committee is required for any trail. Prevent any possibility of misadministration and promptly report to the consentee and the competent authority about misadnistration, adverse reaction or death of a patient, administered with radioactivity A consent is obtained from the patient for all nuclear medicine procedures. The patient should be informed about the safety measures, to avoid radiation exposures to the family members. After administration of radioisotope, the patient should be kept under isolation, if admitted to the hospital. Spread of contamination and radiation exposure to others should be prevented. The nurses and ancillary staff should be instructed about radiation safety and precautions to be adopted during nursing the patient.

Nuclear Medicine Technologist i.

He should ensure proper functioning of all equipments, carry out periodic calibration, QA checks and maintenance ii. He should ensure the purity of radiopharmaceutical, route of administration and the accuracy of dosage before giving to the patient. He should also take precautions to avoid misadministration. iii. Spillage of radioactivity or contamination of the patient, premises, persons and material should be avoided. iv. The RSO should be informed about any mishap in dispensing / administration of dosage to the patient or any unusual incident. He should also assist the RSO in maintaining the records of inventory, use, waste disposal of sources and other safety matters. Radiological Safety Officer (RSO) The RSO assist the consentee to fulfill the regulatory requirements, and ensure safety, security and containment of radioactive sources. He shall carry out radiation and contamination monitoring of work areas, therapy patients, patient areas, management of cadavers containing radionuclides, radioactive waste disposal sites and public areas, and maintain record of findings. The RSO shall ensure that all radiation measuring and monitoring instruments in his custody are properly calibrated and maintained in good condition. He should capable of managing emergency situations, report unusual incident in writing to the consentee and take remedial measures. He should maintain records of the doses of workers, inventory of sources 186 received, used, disposed, any unusual incident, cause of such incident and remedial measures taken.

Regulations and Dose Limits Management of Cadavers Containing Radionuclides 1. All reasonable efforts should be made to remove fluids or organs in which radionuclide is concentrated, provided the collective equivalent dose received in the procedure will be less than that in handling the cadaver as it is. 2. No special precautions are normally necessary for embalming, burial or cremation of the corpse containing the quantities less than that specified in the Table 7.1. Table 7.1: Maximum activities of radionuclides for disposal of corpses without special precautions Radionuclide I Y (colloid) 198 Au (Colliod) 32 P 89 Sr 131 90

Postmortem / embalming, MBq

Burial, MBq

Cremation, MBq

10 200 400 100 50

400 2000 400 2000 2000

400 70 100 30 20

3. The following precautions shall be observed in respect of dead bodies containing quantities more than that given in the above table: i. Cremation: Prior authorization, and specific precautions to be observed during cremation, shall be obtained from RSO. ii. Burial: Relatives shall be prevented from coming into contact with the body and people must not stay near the coffin. The hospital staff and the persons involved in washing, preparing and transporting the body to the burial ground shall be instructed by RSO on dosereducing precautions. The body shall be handled with disposable gloves and kept on plastic sheets to control spread of contamination. iii. Embalming: Embalming is undesirable and, if unavaiodable, shall be done by injection method. All contamination control measures shall be observed under the guidance of RSO. 4. Autopsy on contaminated cadavers shall be performed only in a special autopsy room provided with facilities such as a plastic covered table, water proof flooring, receptacles for organs, instrument cabinet and a washing stand for pathologist. A refrigerator for keeping the cadavers at temperatures below -10 degree shall be provided. The RSO shall supervise autopsy procedures and subsequent decontamination operations. AERB Guidelines for Starting Radioisotope Laboratory Radioisotopes in India can be procured and handled only by the users duly 187 authorized by AERB. This authorization is based on the radiological safety

Textbook of Radiological Safety status of the institution intending to establish a radioisotope laboratory. For this purpose it is mandatory that the plan of the radioisotope laboratory is approved by AERB from radiation safety standpoint. The plan of the radioisotope laboratory will depend upon the type of the radioactive material used, its physical from, activity and the type of experiments to be carried out using the radioactive materials. In general, the minimum requirement is of two rooms of suitable dimensions (adjacent to each other), one for storage and handling of radioisotopes and the other for counting of radioactive samples. Based on the above, one should send to AERB, two copies of the layout of the radioisotope laboratory (drawn to scale 1:50) indicating therein the rooms meant for storage and handling of radioisotopes and counting of radioactive samples. It should also indicate the dimensions of the rooms, positions of the doors, windows, exhausts, fume hoods, workbenches and other fixtures. A site plan (drawn to scale 1:500) of building should also be sent marking clearly therein the location of radioisotope laboratory and the occupancies around it including those above the ceiling and below the floor, if any. The institution is required to pay Rs. 750/- (for Govt. department, University, Public Sector undertaking)/Rs.1500/- (for Pvt. Bodies) as charges towards the approval of the plan. After approval of the plan and equipping and furnishing the laboratory as per this guideline, the institution should approach AERB for commissioning the laboratory for handling radioisotopes and for issuance of authorization / NOC for procurement of radioisotopes. Further, the radioisotope laboratory has to be classified for regular procurement of handling of radioisotopes. The classification will depend upon the quantity of radioisotope used and the required handling facilities. AERB Classification of Radioisotope Laboratories

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The Atomic Energy Regulatory Board (AERB) proposes to classify institution using unsealed radioisotope for non clinical applications in the country and to specify the activities of the radioisotopes which can normally be handled by each user institution. This classification will be broadly based on the relevant recommendations of the International Commission on Radiological Protection (ICRP) and International Atomic Energy Agency (IAEA). Accordingly the radioisotopes have been arranged in four groups based on their radio toxicity as per Annexure-I. Based on the facilities available as per Annexure-II the institution using radioisotope will be divided into three types, namely TYPE-I, TYPE-II and TYPE-III laboratories. The maximum authorized limits of activities that can be procured routinely from Board of Radiation and Isotope Technology (BRIT), will depend on the classification of the laboratories and

Regulations and Dose Limits radionuclides which can be handled at a time will also depend on whether the operations are (i) simple wet, (ii) complex wet, (iii) simple drying, and (iv) dry and dusty. Once the laboratories are classified, the supply of the radionuclides within specified limits will be effected by BRIT, without clearance for individual radionuclides from AERB, in most cases. The recommended procedure for supply of radioactive material is given in the enclosed Annexure-III. The table giving activities of radionuclides of different group that can be procured and handled by each type of laboratories is also enclosed in Annexure-IV. If two or more radionuclides from the same group or different group are ordered, the quantities ordered should be so adjusted that the overall limits are not exceeded. These procedures will help in reducing the delay in the authorization and supply of radionuclides available in India, to a considerable extent. However, this will also involve more detailed accounting of radionuclides by the user institution. AERB Guidelines to Set Up a Radioimmunoassay (RIA) Laboratory Introduction Radioimmunoassasy has been established as a versatile and unique procedure. The advantage of this technique is that it does not involve administration of radioisotopes to the patient and so no radiation exposure to the patient. These assays are useful for the clinical evaluation of the concentration levels of vitally important biological ingredients such as hormones, vitamins, steroids, drugs etc., thereby enabling early diagnosis of various diseases and better management of treatment. RIA work involves the handling and use of very small quantities of radioisotopes, usually not exceeding 3.7 MBq of lodine-125 and Tritium (3H). Radiation Protection Rules (2004) promulgated under the Atomic Energy Act, 1962 requires that the user should obtain authorization from Atomic Energy Regulatory Board (AERB) for handling radioisotopes. Usually, these authorizations are issued subject to the user satisfying the basic safety requirements and adequate trained staff. Minimum Facilities Required The equipment and facilities normally available in a hospital or pathological laboratory can be readily supplemented and used for RIA work. If readyto-use kits are employed, the manipulations are simple and do not result in any significant radiation exposure to the working personnel. Normally no personnel monitoring is required for persons working in RIA laboratories. However, precautions should be taken to safeguard against spread of contamination to the counting tubes and the counter itself. This can be achieved to a large extent by good work practice. Any liquid waste 189

Textbook of Radiological Safety arising from the RIA procedure can be disposed off in the sink provided in the storage area and the used RIA vials should be disposed off in such a way as to avoid reuse. One of the suggested methods is to crush them before disposal. The personnel engaged in the actual work should have adequate knowledge of the basic procedures of counting and should be aware of some simple precautions to be taken in handling of radioactivity. A four weeks training course on “RIA and its Clinical Applications’ is conducted normally twice a year by the Radiopharmaceuticals Division of BARC. For further details, correspondence can be had with Head, Radiopharmaceuticals Division, BARC, Trombay, Bombay 400 085. If labeling with radioiodine (involving use of a few millicuries of 125I) is contemplated, extra precautions and facilities will be required. Procedure for Authorization As a first step the intending user should send two copies of the layout of his proposed RIA laboratory to Head, Radiological safety division, Anushaktinagar, Mumbai-400 094 for approval. The AERB may review the layout, suggest modifications from radiation safety standpoint, which should be implemented by the party and a revised plan submitted for final approval. If found suitable from radiation safety point of view, one of the copies of the layout will be duly approved and sent back by AERB to the institution. When the laboratory is duly constructed and equipped with fittings, fixtures and handling equipment and the qualified staff duly appointed, AERB should be approached for issuance of NOC/authorization for the procurement of RIA kits. If the information is in order and facilities found adequate, AERB will issue the requisite NOC/ authorization. In India, RIA kits are supplied by the Board of Radiation & Isotope Technology, Project House, V.N.Purav Marg, Deonar, Mumbai-400 094 on the basis of the authorization issued by AERB to the user. For import of kits, the intending user has to obtain a ‘No Objection Certificate’ from Head, RSD, AERB. For all information pertaining to availability and use of RIA kits supplied by BRIT, the user should correspond with Head, Radiopharmaceuticals Division, BARC, Bombay 400085. For matters pertaining to planning of RIA laboratory, issuance of NOC / authorization and commissioning / decommissioning the RIA laboratory the user is advised to contact Head, RSD, AERB, Mombay 400 094. REGULATORY CONTROL FOR RADIOTHERAPY EQUIPMENT AND INSTALLATIONS Teletherapy Installation

190

1. Handling of a telegamma therapy source/equipment or Linear accelerator shall be done only in accordance with the terms

Regulations and Dose Limits and conditions of a license granted by the competent authority. Compliance with the safety code is a prerequisite for the issuance of the said license. 2. Telegamma therapy sources and equipment/linear accelerator shall meet the design safety specifications stipulated in the safety code. The manufacturer/vendor must obtain design certification from the competent authority prior to marketing the Linear accelerator / tele gamma therapy equipment. A type approval certificate for the sources/equipment manufactured in the country or a No Objection Certificate (NOC) for the sources/equipment imported into the country, as the case may be. 3. The construction of the teletherapy insatallation shall be undertaken only after obtaining prior approval from the competent authority for the room design and equipment layout from the radiation protection point of view. Any change in the parameters necessitating augmentation of radiation shielding or modification in the approved plan shall be carried out only with the concurrence of the competent authority. 4. The teletherapy installation shall be made available by the licensee for inspection by the competent authority or its representative. 5. Decommissioning and disposal of Linear accelerator/Tele gamma sources/equipments shall be undertaken with prior approval of the competent authority. 6. The employer must ensure that persons handling Teletherapy/tele gamma therapy equipment duly abide by the provisions of the safety code and their further elaboration in various guides issued by the competent authority. He must also ensure that any other measures of safety as the competent authority may stipulate at any time in each individual case are duly implemented without any delay. 7. Telegamma therapy sources shall not be transported in public domain with out prior approval of the competent authority. The transport requirement should comply the transport code AERB/SC/TR-1,1986. 8. Teletherapy/Tele gamma therapy sources/equipment shall be used only in the premises authorized in the license. Sources should not be taken out of the said premises for any purpose without the prior approval of the competent authority. 9. Telegammatherapy/teletherapy equipments shall not be lent, gifted, transferred, sold or disposed off by the licensee with out the prior approval of the competent authority. 10. Any person who contravenes the provisions of these rules or any other terms and conditions of approval granted to him/her by the competent authority, shall be punishable under sections 24, 25 and 26 of the Atomic Energy Act, 1962. The punishment may include fine, imprisonment, or both, depending on the severity of the offence.

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Textbook of Radiological Safety Brachytherapy Sources, Equipment and Installations

192

1. Handling of a Brachytherapy source shall be done only in accordance with the terms and conditions of a license granted by the competent authority. Compliance with the safety code is a prerequisite for the issuance of the said license. 2. Brachytherapy sources and equipment shall meet the design safety specifications stipulated in the safety code. The manufacturer/vendor must obtain design certification from the competent authority prior to marketing the teletherapy / tele gamma therapy equipment. A type approval certificate for the sources /equipment manufactured in the country or a No Objection Certificate (NOC) for the sources/ equipment imported into the country, as the case may be. 3. The construction of the brachytherapy installation rooms shall be undertaken only after obtaining prior approval from the competent authority for the room design and equipment layout from the radiation protection point of view. Any change in the parameters necessitating augmentation of radiation shielding or modification in the approved plan shall be carried out only with the concurrence of the competent authority. 4. The brachytherapy installation and its sources shall be made available by the licensee for inspection by the competent authority or its representative. 5. Decommissioning and disposal of Brachytherapy sources/equipments shall be undertaken with prior approval of the competent authority. 6. The employer must ensure that persons handling teletherapy / tele gamma therapy equipment duly abide by the provisions of the safety code and their further elaboration in various guides issued by the competent authority. He shall also ensure that any other measures of safety as the competent authority may stipulate at any time in each individual case , are promptly implemented. 7. Radioactive sources shall not be transported in public domain with out prior approval of the competent authority. The transport requirement should comply the transport code AERB/SC/TR-1,1986. 8. Radioactive sources shall be used only in the premises authorized in the license. Sources should not be taken out of the said premises for any purpose without the prior approval of the competent authority. 9. Radioactive sources shall not be lent, gifted, transferred, sold or disposed off by the licensee with out the prior approval of the competent authority. 10. Any person who contravenes the provisions of these rules or any other terms and conditions of approval granted to him/her by the competent authority, shall be punishable under sections 24, 25 and 26 of the Atomic Energy Act, 1962.The punishment may include fine, imprisonment, or both, depending on the severity of the offence.

Regulations and Dose Limits Personnel Requirements Every institution having a radiotherapy facility shall have qualified full time Radiation therapist, Medical physicist, Technician (Radiation therapy), Radiological safety Officer (RSO) Level-III and a service engineer. The minimum qualification and experience required are given the safety code AERB/SC/MED-1 and AERB/SC/MED-3. Personnel Responsibilities The team comprising of radiation therapist, Medical physicist and Therapy technicians shall carry out radiation therapy with due regard to patient protection and operational safety in handling the tele-gamma therapy / linear accelerator/Brachytherapy sources and equipment. However, the ultimate responsibility of proper treatment shall vest with the radiation therapist. 1. License: The responsibility of ensuring radiation safety, availability of qualified personnel and providing them requisite facilities to discharge their duties and functions shall rest with the licensee. He shall ensure due compliance with the terms and conditions of the license issued to him by the competent authority. Further he shall provide all necessary facilities to the RSO to discharge his duties and functions. 2. Employer: The employer shall provide adequate number of personnel and facilities to the licensee to discharge his responsibilities effectively. It is the responsibility of the employer to inform the competent authority if the RSO, the licensee or technologists leaves the institution. 3. Manufacturer: The manufacturer /vendor of the equipment shall provide to the user the procedures to QA tests, operating manuals and operation and maintenance details and a copy of safety documents issued by the competent authority from time to time. The vendor shall provide the central axis depth dose and isodose curves for single fields. In the case of Brachytherapy, the vendor must provide the certified strength and out put of each of the sources along with isodose curves. 4. Radiological safety officer (RSO): The RSO shall instruct all radiation workers on relevant safety measures, educate and train new entrants, implement all radiation surveillance measures. He shall control storage and movement of sources in Brachytherapy and conduct periodical surveys. He shall maintain proper records of the personnel doses, conduct periodic radiation surveys and take suitable local measures, including clear administrative instructions in writing, to deal with radiation emergencies. The RSO shall ensure that all radiation measuring and monitoring instruments are properly calibrated and maintained in good working condition. All radiation workers should be trained by the RSO in the management of radiation emergencies.

193

Textbook of Radiological Safety AERB Specification for the Layout of Radiotherapy Facility

194

To establish a Radiotherapy facility, the user institution must go through the Regulatory requirements as mentioned in the Atomic Energy (Radiation Protection) Rules, 2004 and AERB Safety Codes (AERB/SC/MED-1 and 3). No regulatory clearance is issued for establishing the radiotherapy facility by AERB, unless the user complies with the regulatory requirements, specified in these documents. The first step to establish a Radiotherapy facility is to submit the layout plan of the radaition installation and get it approved from AERB from radiation safety standpoint. It is advisable to take services of experienced Radiation Oncologists, Medical Physicists, the Supplier of Radiotherapy Equipments and Architects to prepare the layout plan of the Radiotherapy facility. The room layout plans (to scale 1: 50) and the site layout plan (to scale 1: 500) must be prepared and sent along with the filled in proforma AERB/RSD/RT/PLAN to the Head, Radiological safety Division, Atomic Energy Regulatory Board, Niyamak Bhavan, Anushaktinagar, Mumbai-400094 for approval. i. Plan approval: It is recommended to prepare the layout drawings (to scale) as per the standard layouts prepared by AERB and submitted to the Head, Radiological Safety Division, AERB for approval. The standard layouts are prepared based on typical workload for various facilities and for full occupancies around the installation from radiation safety standpoint. The standard layouts are advantageous as they allow structures to be built in future around the installations without any modification. ii. Site selection: The location of the radiotherapy installation should be so chosen that it is away from unconnected facilities and is close to the related facilities such as Simulator Room, Mould Room, Patient Waiting Area, Treatment Planning System Room, Radiation Oncologist(s) Room, Medical Physicist (s) Room etc. iii. Construction material: The construction material to be used for radiotherapy room should be concrete of density 2.35 gm/cc. However, where structural requirements so demand, RCC may be used. In case hematite concrete is used, the thicknesses may be reduced in inverse proportion to the ratio of the densities. iv. Viewing system: For observing the patient under treatment and the gantry movement from the control room (in case of teletherapy) appropriate viewing system must be provided. This can be achieved by providing Closed Circuit TV System (CCTV). In addition a backup arrangement must be made, which can be achieved by either having a spare Closed Circuit TV System (CCTV) or by using mirrors and an observation window of appropriate dimensions and at a convenient height from the floor must be provided in the interloacked door for observing the patient conveniently from the operator’s position.

Regulations and Dose Limits v. Conduit: A conduit of 5 cm diameter should be provided in the wall as shown in the standard drawings to enable cables of radiation measuring instruments to pass through from the control room to the treatment room. The conduit should be fixed in the specified wall at an angle between 20° to 45° to the horizontal. The lower end of the conduit should be located in the treatment room at a height between 15 cm to 20 cm from the inside finished floor level. Any conduit required in the maze wall can be parallel to the floor at a height of 15 cm to 20 cm from treatment floor level. vi. Door interlock: The door leading to the treatment room may be an ordinary wooden door of width 150 cm. This door should be so interlinked to the control panel by electrical interlocks that the unit cannot be operated when the door is open. The door opening to control room, which is provided to keep the control panel as well as the treatment room secured, may be a collapsible grill door or any other type as per requirement. vii. Air conditioning: The treatment room should be air-conditioned. In case central air-conditioning is to be provided in the radiation therapy room, the ducts for central air conditioning should be taken along the wall of the entrance door and left at the desired location without making any opening on any wall. In case split air conditioners are to be provided in the radiation therapy room, conduits of minimum diameter consistent with the requirement and making an angle between 20° to 45° with the horizontal should be provided. The opening of these conduits in the treatment room should be at a height about 1.5 m from the floor of treatment room. The details may be finalized in consultation with AC Engineers. Altoght it is not common nowadays, if window air-conditioning is to be provided in the radiation therapy room, it must be provided on the secondary walls(not on the primary wall) in case of teletherapy installation. Moreover, the lower end of the openings should be located at a minimum height of 250 cm from the floor level outside and should further be covered with a baffle arrangement. The thickness of the baffle must not be less than 30 cm of concrete and width of the baffle and the length of its vertical portion should be such that 30 cm wide overlap is available all around the openings. The dimensions should be the minimum required for fixing the air conditioners. The length may be decided in consultation with the firm installing the air conditioners and should be such as to permit efficient functioning of the air conditioners. In the case of accelerator installations special ventilation arrangements are required. It is desirable that the control room is also air conditioned. Air conditioners for the control room may be located anywhere in its brick walls as per convenience.

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Textbook of Radiological Safety viii. Warning lights: A red warning light should be provided above the interlocked door and should be so interlinked to the control panel that the light glows when the source is in the ‘ON’ position. ix. Other requirements: Electrical ducting requirements and also any pit, load specifications, conduit etc. should be decided in consultation with the firm installing the unit, before commencement of the actual construction work. x. Construction restraints: It may be necessary for the installation of the unit that some portion of the wall or ceiling be constructed after bringing the crates carrying the unit into the treatment room. This may be decided in consultation with the firm installing the unit. It may also be ensured from the supplier of the unit before starting construction work that the maze/labyrinth provided in the drawing is adequate for the movement of the various components of the radiotherapy unit with or without crates. xi. Associated facility: Associated and supporting facilities of the radiation therapy department include simulator room, treatment planning system room and mould room, doctors room, physicists room, examination room, etc. relative to the radiotherapy room. The institution must show all these facilities in the installation site plan. xii. Ramp: In case of Telecobalt installations, a ramp may be provided in close proximity to the teletherapy installation to facilitate easy movement of the crates carrying the unit to the teletherapy room. The ramp is also useful in future during source replacement operation, which is to be carried out once in every 5-7 years. For this work, a new cobalt-60 source is brought in a transport container weighing 2-3 tons. This container is to be unloaded from a truck and taken into the telecobalt room. The height and slope of the ramp should be so adjusted that the transport container can be unloaded with ease from the truck and transported into the teletherapy room on a suitable trolley. xiii. Beam stopper: In certain instances, it is difficult to meet the shielding requirements of a teletherapy installation due to structural and space constraints. This situation may arise when (i) a teletherapy unit is to be installed in an existing room, (ii) a stationary telecobalt unit in an existing installation is to be replaced by a rotational telecobalt unit or (iii) a telecobalt unit is to be replaced by an accelerator. This difficulty can be circumvented by installing a teletherapy unit with a beam stopper. The beam stopper completely intercepts the primary beam and reduces its intensity approximately by a factor of 1000 and thereby decreases the shielding requirements for the primary barrier. xiv. Starting construction work: No construction work should be undertaken by the institution unless prior approval of AERB for the specific layout of the installation has duly been obtained by the 196 institution.

Regulations and Dose Limits DOSE LIMITS Several scientific groups provide information and recommendations concerning radiation safety. These groups include the National Council on Radiation Protection (NCRP), the International Commission on Radiological Protection (ICRP), the International Atomic Energy Agency (IAEA), and the American National Standards Institute (ANSI). Scientists with these agencies have determined acceptable dose limits for the radiation worker. No clinical evidence of harm would be expected in an adult working within these limits for an entire lifetime. Committees of scientists in the field of radiation science and biology periodically review the literature and, if indicated, recommend changes in the dose limits. These groups provide only recommendations without the force of law and do not enforce or establish radiation safety policy. Dose Philosophy The aim of radiation protection should be to prevent deterministic effects and to limit the probability of stochastic effects to levels deemed to be acceptable. This could be achieved, (i) by setting limits well below threshold dose to deterministic effects, and (ii) the probability of stochastic effects could be reduced by limiting exposures As Low As reasonably achievable (ALARA). In order to minimize the biological effects associated with radiation, dose limits and administrative control levels have been established. As a general approach, three principles designed to control radiation exposure are, ICRP-60 (1990); The Justification principle The Optimization principle The Dose limitation. a. Justification • All exposure either diagnostic or therapeutic shall be under taken only if the benefit gained out weighs the detriment. • No practice shall be adopted unless it produces a net positive benefit. b. Optimization • All exposures which are justified shall be under taken with a minimum possible dose. • Every effort shall be taken to reduce the dose to As Low As Reasonably Achievable (ALARA), taking into account the economic and social considerations. c. Dose limits • The equivalent doses to individuals result in from above practices should be subjected to dose equivalent limits. • These are aimed at ensuring that no individual is expected to radiation risks that are judged to be unacceptable from these practices in normal circumstances. 197

Textbook of Radiological Safety The Tables 7.2 and 7.4 list the ICRD and NCRP dose limits for various applications of radiation, followed by Government of India (AERB) dose limits. Table 7.2: Dose limits (ICRP -60,1990) Application

Occupational, mSv/year

Public, mSv/year

Effective Dose (Based on stochastic effects)

20* (50 mSv annual effective dose limit and 100 mSv in 5 y cumulative effective dose limits)

1 (if needed, higher values provided that the annual average over 5 y does not exceed 1 mSv)

Eye Lens (Based on deterministic effects)

150

15

Skin (skin 100 sq.cm) (Based on deterministic effects)

500

50

Hands, and feet (Based on deterministic effects)

500

50

Fetus

2, after diagnosis

——

*Averaged over any 5 consecutive years. The maximum effective dose limit is 50 mSv/ year Note: 1mSv= 100 mRem Table 7.3: Dose limits,(NCRP-91,1987) Application

Occupational, mSv/year

Public, mSv/year

Effective Dose Eye Lens All others (Skin, extremities, breast,lung, etc.) Embryo-Fetus

50 150 500

1 50 50

5, (0.5 mSv per month)

——

Dose limits (AERB, Government of India, 2001) Workers 1. The cumulative dose over a block of five years shall not exceed 100 mSv 2. The effective dose in any calander year during a five year block shall not exceed 30 mSV. 3. a. The equivalent dose in any calendar year to the lens of the eye shall not exceed 150 mSV. b. The equivalent dose in any calendar year to the skin,the hands and 198 feet shall not exceed 500 mSv.

Regulations and Dose Limits 4. In case of women worker of reproductive age,once pregnancy has been established, the conceptus shall be protected by applying a supplementary equivalent dose limit to the surface of the womens abdomen (lower trunk) of 2 mSv for the remainder of the pregnancy.Internal exposures shall be controlled by limiting intakes of radionuclides to about 1/20 of ALI. The employment shall be of such type that it does not cary a probability of high accidental doses and intakes. Trainess 5. The effective dose in any calendar year shall not exceed 6 mSv. Public 6. The effective dose in any calendar year shall not exceed 1 mSv. 7. In special circumstances, a higher value of effective dose is allowed in a single year, provided that the effective dose averaged over a 5 year period does not exceed 1 mSv/y. Why the Public Dose Limits is less? For a variety of reasons, dose limits for the general population are set lower than those for radiation workers. Justifications for this approach include the following: • The public includes children who might represent a group at increased risk and who may be exposed for their whole lifetime. • It is not the decision or choice of the public that they be exposed. • The public is exposed for their entire lifetime; workers are exposed only during their working lifetime and presumably only while on the job. • The public may receive no direct benefit from the exposure. • The public is already being exposed to risks in their own occupations. • The public is not subject to the selection, supervision, and monitoring afforded radiation workers. Dose Limits to Patients Dose limits do not apply for radiation exposure of patients, since the decision to use radiation is justified depending upon the individual patient situation. When it has been decided that a medical procedure is justified, the procedure should be optimized. This means that the conditions should achieve the clinical purpose with the appropriate dose. Safe limits are determined only for the staff and not for patients. Radiation Limits for Shielding Design NCRP 1993, recommends a fraction of one-half of that effective dose (E) value, or 5 mSv /y, and a weekly shielding design limit of (P) of 0.1 mGy 199

Textbook of Radiological Safety air kerma for controlled areas. In case of uncontrolled areas the effective limit shall not exceed 1 mSv/y. This recommendation can be achieved for the medical radiation facilities with a weekly shielding design limit of 0.02 mGy air kerma for uncontrolled areas. In the United Kingdom, a design dose limit of 6 mSv/y is used for controlled areas. If no special procedures are to be performed, then the dose will be distributed evenly throughout the year and the weekly dose limit will be (6 ÷ 50) 0.12 mSv· week–1. For public areas a design limit of 0.3 mSv/y is used, or (0.3 ÷ 50) 6 μSv· week–1. Depending on local regulations, other limits may be applied and different barrier thickness may be calculated. Annexure-I Classification of Isotope (According to Relative Radio-toxicity per Unit Activity)

Annexure-II Criteria for grading laboratories Type-I (Simple) • A simple chemical laboratory with good ventilation

200 • Two rooms, one for handling and one for counting

Regulations and Dose Limits • • • • • • •

Contamination Monitor Ordinary storage (with security) Sink – ordinary Table surface to be covered with smooth non-absorbent material Remote handling tongs Propipettes / Remote pipettes Foot operated dustbins.

Type-II (Medium) • • • • • • • • • • • • •

Three rooms/more – storage, preparation and one/more handling rooms Special table, floor and wall surfaces Proper ventilation Storage safe – concrete/steel/lead Stainless steel sink (elbow/foot operated tap) Fume-hood with special exhaust system Contamination Monitor & Radiation Surveymeter Personnel Monitoring Badges Planned radioactive waste disposal methods Face mask, Glove box, Surgical gloves Remote handling tongs Propipettes / Remote pipettes Foot operated dustbins.

Type-III (Stringent) Large-scale laboratory – multiroom complex with clear segregation of areas based on use, scale and type of operation with the radioisotopes, the actual facilities required by the user will be determined. A general list is given below • Special table, floor and wall surfaces • Proper ventilation • Storage safe – concrete/steel/lead • Stainless steel sink (elbow/foot operated tap) • Fume-hood with absolute filter incorporated near the junction of hood and ventilation duct • Contamination Monitor and Radiation Surveymeter • Air/Alarm monitor • Foot, Hand and clothing monitor • Pocket monitor • Whole Body Counter • Personnel Monitoring Badges • Bioassay • Dilution and Distribution room • Decontamination room • Respirators

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Textbook of Radiological Safety • • • •

Shoe barrier Master-Slave manipulator Planned radioactive waste disposal methods Foot operated dustbins. Annexure-III Required procedures for expeditious supply of radioactive material for research users handling unshielded sources

1. Specify the nature and quantity of radionuclides available with your department / institution on the date of placing orders for radionuclides with the Senior Manager, Technical Co-ordination & Logistics, Board of Radiation & Isotope Technology (BRIT), V.N.Purav Marg, Deonar, Mumbai – 400 094. 2. Specify clearly the type of operation with radionuclide to be procured (eg. simple dry operation, complex wet operation, normal chemical operation etc.) At the time of ordering for the radionuclide with BRIT/requesting No Objection Certificate (NOC) for importing the radionuclide. 3. All orders/requests for NOC of radionuclides should be routed through the Head of the institution/department, specifying clearly therein the name of the authorised person(s),who will be handling the radionuclides and the department, in which the sources will be handled. 4. The institution shall ensure that all radiation workers in laboratories handling or are likely to get exposed to radiation from radionuclides other than H-3, C-14 and S-35 wear individual personnel monitoring badges. These can be had from the Head, Personnel Monitoring Section, CTCRS, BARC, Anushaktinagar, Mumbai – 400 094. 5. The institution/department is recommended to evolve an efficient record keeping system in respect of radionuclides stored/handled, date of procurement of radioisotope along with activity on that date, purpose of use, mode of disposal, person(s) handling, etc. This shall be made available to the officials of this Division while conducting the radiation protection survey of the laboratory for verification. Annexure-IV Classification of research institutions using unsealed sources Group of

202

Prescribed limit for handling radionuclides

radionuclide *

Type-I

Type-II

Type-III

I II III and IV

≤ 5 µCi ≤ 50 µCi ≤ 500 µCi

≤ 5 mCi ≤ 50 mCi ≤ 500 mCi

> 5 mCi > 50 mCi > 500 mCi

Regulations and Dose Limits Modifying factors according to type of operation Type of operation

Example

Normal chemical operations

Analysis, simple chemical preparations With risk of spills Manipulation of powders and volatile radiioactive compound Grinding

Complex wet operations Simple dry operations Dry and dusty operations

Modifying factor 1.00 0.10 0.10 0.01

BIBLIOGRAPHY 1. AERB safety code: Brachytherapy sources equipment and installations, AERB/ SC/MED-3. 2. AERB safety code: Medical Diagnostic X-ray equipment and installations: AERB/SC/MED-2 (Rev.1). 3. AERB safety code: Nuclear medicine facilities, AERB/SC/MED-4(Rev.1). 4. AERB safet0y code: Telegamma therapy equipment and installations, AERB/ SC/MED-1. 5. Atomic energy (Radiation protection) Rules, 2004:Published in the Gazette of India: September 11, 2004. 6. International commission on radiological protection,1990 Recommendations of the ICRP, Publication 60, Pergamon press, Oxford (1991). 7. National radiological protection Board, Doses to patient from medical x-ray examinations in the UK: 2000 review, NRBP-W14, Chilton (2002), 1.

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Chapter

8

Personnel Protection

DIAGNOSTIC RADIOLOGY The personnel working in the radiological departments should receive exposures well below the regulatory limits on the lines of ALARA principle. Hence the training, equipment and work practices are so designed to minimize the time with radiation sources. It should also maximize the distances between the worker and radiation sources, and provide the use of appropriate shielding when working with radiation sources, including the patient. Hence, sufficient protective devices should be offered to the personnel. In addition, there are certain personnel requirements, responsibilities, to be adhered to achieve complete safety. This safety should cover all the personnel including staff, patient and public. RADIOGRAPHY Protective Devices Personnel protection may be achieved by using several protective devices and adopting good work practices. These devices should include lead apron, thyroid shield, gonad shield and lead glass, ceiling mounted barriers etc. especially in radiography and fluoroscopy imaging. All individuals working in the radiation room must wear a lead apron, when the X-ray tube is operated. Lead Apron The lead apron should have a lead equivalent thickness of 0.25-0.5 mm. Usually it is made up of rubber material to provide flexibility and handling (Fig. 8.1). Aprons protect the torso of the body and are available wrap around designs. This is to protect the back side when the worker is exposed to scattered radiation. The attenuation offered by the lead apron is as follows: • 0.25 mm: >90% scattered radiation is attenuated

Fig. 8.1: Lead apron

Personnel Protection • 0.5 mm: 95-99% scattered radiation is attenuated. Though higher thickness aprons offer greater protection, it weighs 50100% more than the 0.25 mm thickness. It is a great matter of concern in fluoroscopy. The weight of the apron become the limiting factor in the ability of the radiologist and other workers to complete the examination. There are some design(Skirt-vest) which put much weight on the hips instead of the shoulders. The lead aprons do not cover the arms, lower legs, the head and neck, thyroid and eyes. Thyroid Shield and Lead Glass Personnel can wear thyroid shield and lead glasses for protection in the imaging room. The thyroid shield (Fig. 8.2A) is made up of lead and wraps around the persons neck. It offer protection similar to that of lead apron. Lead glasses attenuate the X-rays about 30-70%, depending upon the content of the lead. Hence, the weight is a major concern in the lead glass. Normal, lead glasses used in the hospital may offer 20% attenuation. Protective gloves made up of 0.5 mm lead thickness, may offer protection to the hands (Fig. 8.2B).

A

B B

Figs 8.2A and B: Thyroid shield and protective gloves

Ceiling Mounted Barriers Ceiling mounted barriers are used in cardiac catheterization laboratories and interventional imaging works. These devices are placed between patient and the personnel in the room. The ceiling mounted system is counter balanced and easily positioned. Lead glass or leaded acrylic shields are transparent and often provide greater attenuation than lead aprons. Organ Shield Whenever required, suitable shielding material should be used to shield the organs of interest or critical organs. For example when limb (hand) is 205

Textbook of Radiological Safety X-rayed lead apron may be provided to the patient. Similarly gonadal shield (Fig. 8.3)can be provided to the patient to protect the gonads from primary beam. The gonad shield should have a lead thickness of 0.5 mm of lead. Shields should not interfere with the anatomy under investigation. The use of gonad shield can reduce the absorbed dose in the testes by up to 95%, while the reduction of absorbed in the Fig. 8.3: Gonad shield ovaries, can be about 50 %. The eyes should be shielded for X-ray examinations involving high absorbed doses in the eyes, such as conventional petrous bone tomography. This is very important when multiple X-ray examinations are needed. Absorbed dose in the eyes can be reduced by 50-75 % by shielding the eyes. The use of the posterior-anterior projection rather than the anteriorposterior projection can reduce the absorbed dose in the eyes by 95 %. Work Practice Occupancy in the Room Only persons whose presence is necessary should be in the imaging room during the exposure. Overcrowding should be avoided. All such persons must be protected with lead aprons/shields. The X-ray room shall be kept closed during the radiation exposure. Assistance to Patients Holding of children or infirm patients for X-ray examination shall be done only by an adult relative or escort of the patient. Hospital personnel should not hold the patients during imaging procedure. No person should routinely hold patients during diagnostic examinations, certainly not those who are pregnant or under the age of 18 years. Such persons shall be provided with protective aprons and gloves. Immobilization devices (Mechanical supporting or restraining devices) shall be used to prevent movement of children during exposure. In no case shall the film of X-ray tube be held by hand. In no instance shall the holder’s body be in the useful beam, and should be as far away from the primary beam as possible. Protection in Radiography The goal of diagnostic radiology is to have optimal balance between image quality and dose to the patient. Any request for X-ray imaging needs to be justified, on benefit vs risk point of view. When complex examinations involving X-rays are referred, one has to find the truthfulness of the 206 reference. There is a need for standardization of techniques and procedures and optimization of protection measures.

Personnel Protection Equipment and Peripherals Patient dose can be reduced by selecting optimal equipment and its peripherals. Constant potential / High frequency generator can reduce dose significantly. The other methods are use of fast screen-film combination (e.g. rare earth, 400 speed for pediatric), low attenuation (e.g. carbon fiber) materials for cassette fronts, antiscatter grid interspacing and table tops and grid removal etc. Tube Voltage (kVp) Increasing kVp result in greater transmission of X-rays through the patient. This means that the absorption is lesser in the patient even though the exposure / mAs is higher. The best way to reduce patient dose is reduction of mAs. But this will reduce the image contrast due to higher effective energy of the X-ray beam. Hence, one has to balance between patient dose and image contrast. As a rule of thumb, patient exposure can be reduced by using a higher kVp and lower mAs. Filtration A filter is a metallic sheet introduced in the path of X-rays in order to reduce patient dose. Diagnostic X-ray consists of both low and high energy Xrays. The high energy X-rays transmit through the patient and contribute to the image formation. Whereas the low energy X-rays are absorbed in the first few cm of tissue, there by increasing the radiation dose to skin. Filtration can remove selectively low energy X-rays, and there by reducing patient dose and skin injuries. As the filtration is increased, the beam become hardened and decreases the image contrast. Filtration also decreases tube output and hence an optimal filtration is required for each X-ray unit. The recommended beam filtration is follows: i. General radiography • 1.5 mm Al below 70 kVp • 2.0 mm Al between 70 and 100 kVp • 2.5 mm Al above 100 kVp. ii. Mammography • Be 1 mm + Mo 0.03 mm (Mo target) • Be 1 mm + Rh 0.025 mm (Rh target). In mammography, Mo target with Rh filter is commonly used. Whereas Mo can not be used as filter in mammography X-ray tubes with Rh targets. Field Area Reduction of the field size by collimation is another important dose reduction technique. Hence, minimal field size to cover the patient volume is sufficient. Field size reduction reduces the scatter there by reducing the dose to adjacent organs. The scatter incident on the detector also decreases, 207

Textbook of Radiological Safety resulting in improved image contrast. Hence, the rule of thumb is “use smallest possible field size and good collimation”(Fig. 8.4A and 8.4B).

208

Figs 8.4A and B: Field size and dose reduction

Personnel Protection The decrease in X-ray field size also reduces the total radiation energy delivered to the patient, and therefore the mass of the skin and internal tissues irradiated. In radiography projections, the gonads should be kept outside the X-ray beam by carefully adjusting the X-ray field. When the testes are located just a few centimeters outside the X-ray field edge, the absorbed dose in the testes can be one-fourth or less of that when testes are in the field (Fig. 8.5).

Fig. 8.5: Change in absorbed dose in testes with distance between edge of the X-ray field and testes

Source to Object Distance The radiation intensity from a point source varies inversely as the square of the distance from the source. Increasing the source to object distance (SOD) and source to image distance (SID), will reduce the patient dose. As the SOD/SID is increased there is a reduced beam divergence, which reduce the volume of patient irradiation. This will enable us to decrease the integral dose. Increased SOD also facilitate reduction of patient exposure due to tube leakage, since the tube is away from the patient. Where as decrease of SOD, increases the radiation intensity sharply at the surface of the patient (Fig. 8.6). In radiography and fluoroscopy with stationary X-ray equipment, the SOD should be not less than 45 cm. When the SID is less than 100 cm, the quality of the diagnostic information becomes poorer, there fore longer SID has clinical advantages. Photofluorography and radiography of the chest should be performed with a SID of at least 120 cm. In the case of C-arm units fixed SID is used, therefore increase of SOD is the only way of 209

Textbook of Radiological Safety reducing patient dose. Hence in fluoroscopy minimum distance between source and the patient must be not less than 30 cm.

Fig. 8.6: Dependence of absorbed dose in the skin on the distance from the X-ray source; the skin to image receptor distance is constant at 25 cm

Carbon Fiber Materials The use of carbon fiber materials for the patient support, in anti scatter grids and for the radiographic cassette face, in place of conventional materials, allows transmission of a larger proportion of the X-ray beam. At an X-ray tube voltage of 80 kV, the use of carbon fiber materials enables the absorbed dose in the skin of the patient to be reduced. The overall reduction of absorbed dose in the skin of the patient facing the X-ray tubes, from the combined use of carbon fibre materials in patient supports, anti scatter grids and radiographic cassettes, is in the range of about 30 % to more than 50%. If the X-ray tube voltage is not changed, the percentage reductions in absorbed dose in deeper tissues will be similar. Image Receptors The speed of the image receptor determines the number of X-ray photons necessary to produce an optimal image signal. This is directly related to the patient dose. Higher speed (400) system require less exposure to produce the same optical density, and decreases the patient dose. Faster film increases the quantum mottle and faster screen (thick) decreases the spatial 210 resolution. Thus, higher speed film-screen reduces the patient dose, but limited by image quality.

Personnel Protection Computed radiography (CR) uses photostimulable storage phosphor (PSP) detectors and the digital radiography (DR) receptors have wide latitude. These systems allow post processing methods, to manipulate image density for optimal viewing. They compensate for under or over exposure and reduce retakes. Under exposed images are associated with high quantum mottle and poor contrast resolution. Over exposed images contribute higher patient dose. The PSP receptor system is equivalent to 200 speed screen-film system in terms of quantum mottle and noise for adult abdomen and chest radiography. In the case of extremities imaging, the CR should be used at higher exposure levels (e.g. speed 75-100). Whereas in pediatric imaging increased speeds (e.g. 400) is recommended. Conventional screen-film has inherent safety feature. If high kV is used the film becomes over dark, there by increasing the patient dose. One can easily identify this and carry out necessary remedial measures. In digital image receptors, one can not notice this adverse effect on image quality, since it is adjusted in post processing. Therefore, higher patient exposures may go unnoticed. Hence, digital image receptors require strict quality assurance check. Image Intensifier The image intensifier has wide dynamic range and the entrance exposure is controlled by light limiting aperture or electronic of the subsequent detector (e.g. TV camera).The entrance exposure is 1mR per image in fluoroscopy, and 4 mR per image for digital subtraction angiography. Any reduction in exposure is limited by quantum mottle. Patient Positioning The collimator is adjusted to exclude radiosensitive organs such as gonads, breast and eyes (Fig. 8.7). Patient Motion Patient motion is a matter of concern in diagnostic imaging. It may cause motion artifacts, which may increase repeat X-rays. This will increase the patient dose. To reduce patient motion (i) short exposure times, (ii) use of immobilization or sedation, (iii) entertainment, or distracting devices should be applied and adopted. In addition, reduction of repeat films can also reduce patient dose significantly. Hence, the film reject rate due to all causes should be kept below 5%. Proper instruction to the patient, checking the factors before exposure, proper darkroom procedure, and periodic maintenance of automatic processors may also help to reduce repeat X-rays and patient dose.

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Textbook of Radiological Safety

Figs 8.7A and B: (A) Wrong positioning, and (B) correct positioning

PROTECTION IN FLUOROSCOPY Fluoroscopy imaging contributes large portion of dose in medical imaging due to continuous X-ray production and real time image output. Though the exposure technique are moderate (3 mA, 80 kVp), the examination on time extend from minute to hours. Cini angiography studies employ high exposure rates of 20 to 70 R /min with short exposure times. Some systems have turbo mode where the exposure rate may exceeds 20 R/min and hence this mode should be used judiciously. The patient dose also depends upon the angulation of the beam through the patient. Exposure rates are greater in lateral examination than that of anteroposterior. For example, a fluoroscopy imaging involves a technique of, 2 mA, 80 kVp,10 min on time, delivers an exposure of 6 R at 1 m. Then the skin entrance exposure will be: 2

⎛ 100 ⎞ Entrance exposure = ⎜ ⎟ × 6 = 67 R ⎝ 30 ⎠

Thus, in fluoroscopy the entrance dose is higher. The entrance exposure of various imaging systems like Radiography, fluoroscopy and CT scan are given in Table 8.1. Table 8.1: Radiation dose in various imaging systems

212

EXAM

Radiography (mGy)

Fluoroscopy (mGy)

CT (mGy)

Chest Abdomen / pelvis

0.14 0.53

3.6 6.4

13.2 21.5

Personnel Protection The source to object distance should be not less than 30 cm. The patient entrance dose is limited to a maximum of 10 R / min, with automatic exposure control (AEC), and 5 R / min without AEC. The X-ray tube is provided with multiple focal spots (0.3, 0.6, and 1.2 mm) to handle adult and pediatric patients. Intensifiers are available in different sizes (4.5, 9, and 12 inch) and proper selection of intensifier mage size to match a specific institution is very important. Dose Reduction in Fluoroscopy The most important method of dose reduction in fluoroscopy is to limit the beam ON time, by using short burst of exposures. One can also use the last image-hold facility, to view the image even after the exposure. This will reduce the fluoro time by 50-80% in many procedures. Pulsed fluoroscopy reduces dose by allowing frame rates lesser than real time. It may be used along with digital image memory to reduce patient dose. This is a suitable technique in patients where higher temporal resolution is not required. Use of optimal image intensifier system (conversion gain, high contrast ratio, spatial resolution and contrast sensitivity) can also reduce the patient dose. Reducing the field size by collimation will also reduce the integral dose. Maintaining much distance between the X-ray tube and patient will also reduce the skin dose. Though magnification technique improves spatial resolution, it also increases the patient dose and hence used sparingly. Dose -Area –Product (DAP) meter should be used to measure the dose in fluoroscopy. It is a radiolucent ionization chamber positioned across the primary beam, beyond the collimator (Fig. 8.8). The ionization signal generated is proportional to the beam area. If the DAP meter is calibrated with field size, source to object distance (SOD), source to detector distance (SDD) and collimation, then patient dose can be estimated. Real time display of patient dose can help the physician to modify the procedure, to minimize skin injuries. It is easy to identify the patient who may be at risk, due to radiation. Hence, physician training plays an important role in dose reduction methods.

Fig. 8.8: Dose area product meter

213

Textbook of Radiological Safety PROTECTION IN COMPUTED TOMOGRAPHY Computed tomography (CT) is a valuable and life saving imaging tool in medical imaging. In CT scans automatic exposure modes are not available. Most CT scanners use fixed kVp and mAs regardless of patient size. Fixed techniques in CT may lead to unnecessary patient dose; e.g. neonatal, children. CT scan is the major contributor of patient dose. It performs 8 % diagnostic issues, but provides 48% of the population’s collective dose (US data) as shown in the Fig. 8.9. Few models now offer modulation of mA as a function of rotation angle. This will compensate the thickness variation in torso of body from AP to lateral projections, which will reduce the patient dose. Therefore, the training of the technologists is very important in CT scan imaging. He should know the way of reducing patient dose, by reducing mAs for thinner patients.

Fig. 8.9: Frequency distribution of medical imaging and patient radiation dose in %

The Food and drug administration (FDA), USA has issued a public health notification (2002) regarding CT scan dose reduction that includes: i. Reduce tube current, (increases the image noise), ii. Increase pitch or table increment (increase the effective slice thickness), 214 iii. Use noise reduction algorithms,

Personnel Protection iv. Pediatric protocols, v. Develop and use a chart or table of tube-current settings based on patient weight or diameter and anatomical region of interest, vi. Reduce the number of multiple scans with contrast material, vii. Eliminate inappropriate referral. The reduction of current and increased pitch though reduces the patient dose, but increases the noise and slice thickness. Hence, optimal balancing is necessary in devising the techniques. A technical chart can be devised involving kVp, mA and thickness of the patient (Table 8.2). The chart is devised with 120 kVp, 300 mAs, 5 mm slice and pitch 1, for creating a good quality image of a 32 cm torso. For a 30 cm patient about half of the mAs (160) is sufficient to produce the same quality of image. For a 20 cm patient only 7 mAs is required, which reduces the patient dose by 43 %. Thus, the technical chart is very useful to reduce patient dose, with proper selection of mAs for each patient. PROTECTION IN PEDIATRIC IMAGING Children have higher radiation sensitivity than adults and have a longer life expectancy (larger window to express radiation damage). Increasing numbers of radiological examinations are being performed in infants and children. Millions of children undergo high dose procedures such as computed tomography and interventional procedures. A pediatric radiological procedure should be individually planned and projections should be limited to what is absolutely necessary for a diagnosis. Therefore, imaging techniques that do not use ionizing radiation should always be considered as an alternative. Table 8.2: Technique chart (120 kVp, 300 mAs, 5 mm slice, Pitch = 1) Patient diameter (cm)

% mAs

mAs

14 16 18 20 22 24 26 30 32 34 36

0.3 0.6 1.2 2.2 4.2 7.9 15 53 100 188 352

1 2 4 7 (43% dose!) 13 24 45 160 300 564 1,058

Children are more vulnerable to the late somatic effects and genetic effects of radiation than adults (epidemiologic study). Children receive a higher dose, when adult settings are used. Children are 10 times more sensitive to 215

Textbook of Radiological Safety radiation than adults and girls are more sensitive than boys. The Risk for developing a radiation–related cancer can be several times higher for a young child compared with an adult exposed to an identical technique. Radiation Risks in children is a public health issue. Hence, one has to examine the following issues before carry out the imaging: i. Justification of requested examinations, ii. Vetting of referrals for complex examinations, iii. Standardization of techniques and procedures, iv. Optimization of protection measures. Typical values of Entrance Surface Dose (ESD) per radiograph and Dose Area Product (DAP) for common paediatric fluoroscopy examinations are given in Table 8.3. Pediatric Radiography Equipment and Peripherals The selection of equipment associated peripherals should suit the pediatric imaging work, so that the radiation dose will be lesser. This includes (i) fast screen-film combination (e.g. rare earth, 400 speed for pediatric), (ii) low attenuation (e.g. carbon fiber) materials for cassette fronts, anti-scatter grid interspacing, table tops, (iii) grid removal, and (iv) constant potential/High frequency generator. Table 8.3: Typical dose levels in paediatric radiology Examination

Entrance surface dose (µGy) Age

Abdomen AP Chest PA/AP Pelvis AP Skull AP Skull LAT

0

1

5

10

15

110 60 170 / /

340 80 350 600 340

590 110 510 1250 580

860 70 650 / /

2010 110 1300 / /

Dose area product (mGy•cm2) MCU Barium meal Barium swallow

430 760 560

810 1610 1150

940 1620 1010

1640 3190 2400

3410 5670 3170

Special Considerations Dose to the children can be reduced significantly by adopting either one or combination of the following: This includes (i) Higher kVp and lower mAs, (ii) Increased filtration, (iii) Field size reduction (Collimation), (iv) Shielding 216 of gonads, thyroid and lens, (v) Increased Source-Object Distance (SOD), and (vi) Posteroanterior projections in female patients.

Personnel Protection Patient Motion Patient motion should be avoided during examinations under radiation. To fulfill this one can adopt the following: (i) Short exposure times, (ii) Immobilization or sedation of the patient (Fig. 8.10), and (iii) incorporation of entertainment, or distracting devices.

Figs 8.10A and B: Immobilization of child during chest radiography (For color version see plate 3)

Dose Reduction Methods in Pediatric Radiography Anti-scatter grids are normally not required in pediatric radiography as the gain in image quality does not justify the increase in patient dose, except in children in their teens and when the body build is such as to increase scatter • Good image detail is achieved by maintaining a balance between the use of a small focal spot size and a short exposure time. • High speed screen-film combinations should be used where possible to enable reduction in radiation exposure and exposure time, as the reduced resolution obtained is comparatively insignificant for the majority of clinical indications. • The use of Automatic Exposure Control (AEC) is generally not appropriate in children as the sensors (size and geometry) are normally designed for adult patients. Instead, exposure charts corresponding to radiographic technique, patient thickness in the X ray beam and presence or absence of anti-scatter grid are much safer and easier to use. • The radiation beam should be limited using collimation. • Shielding devices should be appropriately positioned to be efficient for protecting the tissues for which they are placed and to avoid unnecessary repeat examinations. • Immobilization, when required, should be provided by specialized 217 devices, if possible.

Textbook of Radiological Safety Pediatric Fluoroscopy Fluoroscopy offer much higher dose to the children than radiography imaging. Hence, optimal selection of equipments and specialized techniques are very much essential. The Image intensifier should have diameter of 4.5 in (11 cm).The generator should have range of mAs from 0.1 to 6, so that one can have varying techniques from 100 mA × 1ms to 800 mA × 7ms. This is suitable for imaging children having weight from 3-140 kg at 65-75 kVp. The X-ray tube should have at least two focal spots namely 0.3 mm (3-4 years child) and 0.6 mm (8 years children). The cine frame rate should be >60 fps. i. Dose reduction methods in fluoroscopy. ii. The patient should be positioned as close as possible to the image intensifier. iii. The X-ray tube should be as far away as possible from the patient table in order to avoid excessive skin dose. iv. The lowest frame rate acceptable and last-image-hold facility should be used. v. Some centres prefer to set a ‘floor’ (a kVp) below which the system will not go, such as 70 kVp for paediatric patients and 80 kVp for adults. vi. Additional copper filtration also reduces patient dose. Pediatric Computed Tomography CT and interventional procedures are high dose procedures in radiology and yield higher individual patient doses than other radiological procedures do. The patient dose in CT is an important issue for children as reports suggest that in some centres the exposure factors used for scanning children are the same as for adults. This problem is relatively lesser in interventional procedures as the machine, on the basis of the body thickness falling in the X-ray beam, automatically adjusts factors in most modern equipment. CT scanning contributes most collective dose from radiographic exposures due to the increasing use of this modality. CT scans are increasingly used in pediatric imaging and mostly fixed kVp and mAs, regardless of patient size is used. There is a potential increase in the radiation exposure to children undergoing these scans. About 4.7 million CT examinations are performed annually on children in the US. CT in children has increased about 8 fold since 1980, with annual growth of about 10 % per year. Children receive higher doses than necessary, when adult settings are used. The effective dose from single pediatric CT ranges from 1-30 mSv, and one third of the children have three scans, which will triple the cancer risk. There is no need for these large doses and CT settings can be reduced significantly with out losing the image quality. Hence, 218 children should not be scanned using adult techniques, and pediatric CT protocol or dose reduction methods should be made available.

Personnel Protection Dose Reduction Methods Using the PMMA phantom (Fig 8.11), ACR CT Accreditation program, the CTDIvol can be calculated. For a adult body this is about 25 mGy and head it is 75 mGy. Taking these as baseline data, dose reduction factors can be designed. The pediatric protocol is obtained as follows: Table 8.4 and 8.5 presents the reduction factors for pediatric abdomen and thorax and head respectively. Pediatric mAs = Baseline data × Reduction factor (RF)

Fig. 8.11: ACR accredited PMMA Phantom (For color version see plate 3) Table 8.4: Patient thickness, age and reduction factors for pediatric abdomen and thorax PA Thick (cm)

Age

Abdomen RF

Thorax RF

9 12 14 16 19 22 25 31

Newborn 1 5 10 15 S Adult M Adult L Adult

0.43 0.51 0.59 0.66 0.76 0.90 Baseline 1.25

0.42 0.49 0.57 0.64 0.74 0.82 Basline 1.16

Table 8.5: Patient thickness, age and reduction factors for pediatric head PA Thick (cm)

Age

Pediatric Head mAs RF

12 16 17 19

New born 1 5 M Adult

0.74 0.86 0.93 Baseline

Example 1: An adult thorax is examined at a technique of 120 kVp, 0.5 sec scan time, 200 mA, pitch=1and FOV=35 cm. What is the appropriate 219 pediatric technique for a 5 year old thorax at a pitch of 1?

Textbook of Radiological Safety From the Table 8.4 the RF for 5 year old thorax is 0.57, then Pediatric mA = Baseline × RF = 200 mA x 0.57 = 144 mA. Example 2: An Adult head is examined at a CT technique of 140 kVp, 0.5 sec scan time, 400 mA, pitch=1 and FOV=25 cm. What is the appropriate technique for a one year old head? From Table 8.5 the RF for a one year old head is 0.86, then Pediatric mA = Baseline × RF = 400 mA × 0.86 = 344 mA. In general a head examination with adult protocol (200 mAs) may give 23-49 mGy organ dose in brain. If the mAs is adjusted to 100 and dose become 11-25 mGy. Similarly the abdomen dose reduces from 21-43 mGy to 5-11 mGy if the mAs is adjusted as shown in the Table 8.6 Table 8.6: Pediatric organ and effective doses with normal and adjusted mAs for head and abdomen examinations Exam

Organ

Organ dose (mGy)

Efective dose (mGy)

Head (200 mAs) Head (100 mAs, adjusted) Abdomen (200 mAs) Abdomen (50 mAs adjusted)

Brain Brain Stomach Stomach

23-49 11-25 21-43 5-11

1.8-3.8 0.9-1.9 11-24 3-6

Dose Reduction Methods in Paediatric Chest CT i. Image quality in CT is generally more than what is required for confident diagnosis. Awareness on this can help in significant reduction in patient dose. ii. Radiologists and physicians should be aware that images with low noise, even if they do not look very crisp, may provide the diagnostic information. iii. mAs reduction at defined kVp has been used with success by many centres and is the most efficient method of dose management in children as also in adults. There is lack of consensus on kVp reduction in CT examination. iv. Many authors suggest using 100 to 200 mAs settings for high resolution chest CT in children. However, reliable diagnostic studies can be obtained using much lower mAs. In cooperative children who are able to breath-hold as low as 34 mAs can be used and in non-cooperative 220 children 50 mAs.

Personnel Protection v. Whenever radiosensitive tissues such as breast and thyroid fall within the exposed area, they should be shielded. Breast-anlage (primordium or the first rudiment of the breast, the underdeveloped tissue) protection using for example 2 mm thick bismuth coated latex shielding reduces the dose to the breast-anlage by approximately 40%. vi. Recent technology developments include automatic tube current modulation where the tube current is adjusted according to thickness and density of tissues to maintain a constant level of image noise. Dose Reduction Methods in Pediatric Abdominal CT Strategies should include obtaining only necessary CT examinations. MRI and US should take priority. If possible, the examination should be tailored to answer the specific question asked by the referring clinician, for example, pelvic scanning is not always necessary when an abdominal scan is requested and it may be possible to curtail follow-up CT exams to a specific organ. In addition, imaging parameters such as kVp and mAs need to be adjusted for patient size. Size-based tables for abdominal multidetector CT and body CT angiography in children are available. In one study, children were classified by colors based on weight and this was shown to significantly reduce scanning errors in settings for pediatric multi-detector CT. Recent technology developments include automatic tube current modulation where the tube current is adjusted according to thickness and density of tissues to maintain a constant level of image noise. Finally, the use of multiphase scanning should be curtailed as much as possible. PREGNANCY AND RADIATION It is unlikely that radiation from diagnostic radiological examinations will result in any deleterious effects on the child, but the possibility of a radiationinduced effect cannot be entirely ruled out. The effects of exposure to radiation on the conceptus depend on the time of exposure with respect to the date of conception and the amount of absorbed dose. Effects on Radiation Exposure In Utero (ICRP-84) i. Prenatal doses from most properly done diagnostic procedures present no measurable increase in the risk of prenatal death, malformation, or the impairment of mental development over the background incidence of these entities. Higher doses, such as those involved in therapeutic procedures, can, however, result in significant fetal harm. ii. There are radiation-related risks throughout pregnancy that are related to the stage of pregnancy and the foetal absorbed dose. Radiation risks are most significant during organogenesis and the early foetal period, somewhat less in the second trimester, and least in the third trimester. 221

Textbook of Radiological Safety iii. During the period of ± 25 weeks post conception, the central nervous system (CNS) is particularly sensitive to radiation. Fetal doses in excess of about 100 mGy may result in a verifiable decrease of IQ. During the same time, foetal doses in the range of 1,000 mGy (1 Gy) result in a high probability of severe mental retardation. The sensitivity is highest 8±15 weeks post conception. The CNS is less sensitive to these effects at 16±25 weeks of gestational age and rather resistant after that. iv. Radiation has been shown to cause leukemia and many types of cancer in both adults and children. Throughout most of pregnancy, the embryo/ fetus is assumed to be at about the same risk for potential carcinogenic effects of radiation as are children. Chest and Extremity Radiography in Pregnancy Medically indicated diagnostic studies remote from the fetus (e.g. radiographs of the chest or extremities) can be safely done at any time during pregnancy if the equipment is in proper working order. Commonly, the risk of not making the diagnosis is greater than the radiation risk involved. If an examination is typically at the high end of the diagnostic dose range and the fetus is in or near the radiation beam or source, care should be taken to minimize the dose to the fetus while still making the diagnosis. Tailoring the examination and examining each radiograph as it is taken until the diagnosis is achieved and then terminating the procedure can do this. CT and Pregnancy Occasionally, a patient will not be aware of a pregnancy at the time of an X-ray examination, and will naturally be very concerned when the pregnancy becomes known. In such cases, the radiation dose to the fetus/conceptus should be estimated, but only by a medical physicist/ radiation safety specialist experienced in dosimetry. The patient can then be better advised as to the potential risk involved. In many cases there is little risk, as the irradiation will have occurred in the first 3 weeks following conception. In a few cases the conceptus will be older and the dose involved may be considerable. It is, however, extremely rare for the dose to be high enough to warrant advising the patient to consider terminating the pregnancy. If a calculation of radiation dose is required in order to advise the patient, the radiographic factors should be noted if known. Some assumptions may be made in the dosimetry, but it is best to use actual data. The patient’s date of conception or date of LMP (last menstrual period) should also be 222 determined.

Personnel Protection Cardiac Catheterization and Pregnancy There will be many situations where the benefit of performing the procedure is much greater than any small possible harm that might arise from the radiation exposure. However, as always with any medical exposure, each particular procedure must be clinically justified, including in this situation taking into account when the procedure needs to occur. Once justified, due care is taken to optimise how the procedure is performed so as to minimise radiation exposure to the fetus, consistent with achieving the desired clinical outcome. The radiation exposure to foetus predominantly arises from scattered radiation within the patient. Some of the main methods for minimizing the dose to the foetus include: i. Restricting the X-ray beam size to being as small as is necessary for the clinical purpose; ii. Choosing the direction of the primary beam so that it is as far away from the foetus as possible; iii. Selecting appropriate exposure factors; and iv. Ensuring that the overall exposure time is as small as possible. For well performed procedures, estimated foetal doses are typically quite small, and well below the level of concern for radiation effects. As a final comment, putting a lead apron on the table to cut down any primary beam from the X-ray tube reaching the fetus has very little effect, but it can be reassuring to the patient and staff and thus is recommended provided the use of the apron does not compromise the performance of the procedure. Termination of Pregnancy after Radiation Exposure According to ICRP 84, termination of pregnancy at fetal doses of less than 100 mGy is not justified based upon radiation risk. At fetal doses between 100 and 500 mGy, the decision should be based upon the individual circumstances. The issue of pregnancy termination is undoubtedly managed differently around the world. It is complicated by individual ethical, moral, and religious beliefs as well as perhaps being subject to laws or regulations at a local or national level. This complicated issue involves much more than radiation protection considerations and require the provision of counseling for the patient and her partner. At foetal doses in excess of 500 mGy, there can be significant foetal damage, the magnitude and type of which is a function of dose and stage of pregnancy. Many believe that this dose can cause sterility in the exposed individual, but really it is not so. The gonads are radiosensitive organs in the human body. The threshold radiation dose for permanent sterility in men is 3500 - 6000 mGy, and for women 2500 - 6000 mGy. As diagnostic X-ray examinations involve small doses (Table 8.7), they imply no risk of sterility. 223

Textbook of Radiological Safety Table 8.7: Approximate foetal doses from common diagnostic procedures in United kingdom (Sharp, Shrimpton, and Buiy, 1998) Conventional X-ray examinations

Mean (mGy)

Maximum (mGy)

Abdomen Chest Intravenous urogram Lumbar spine Pelvis Skull Thoracic spine Fluoroscopic examinations Barium meal (UGI) Barium enema Computed tomography Abdomen Chest Head Lumbar spine Pelvis

1.4 < 0.01 1.7 1.7 1.1 < 0.01 < 0.01 Mean (mGy) 1.1 6.8 Mean (mGy) 8.0 0.06 < 0.005 2.4 25

4.2 < 0.01 10 10 4 < 0.01 < 0.01 Maximum (mGy) 5.8 24 Maximum (mGy) 49 0.96 < 0.005 8.6 79

Continuation of Work of a Pregnant Employee in X-ray Department A pregnant worker can continue working in an X-ray department as long as there is reasonable assurance that the foetal dose can be kept below 1 mGy during the pregnancy. In interpreting this recommendation, it is important to ensure that pregnant women are not subjected to unnecessary discrimination. There are responsibilities for both the worker and the employer. The first responsibility for the protection of the conceptus lies with the woman herself, who should declare her pregnancy to management as soon as the condition is confirmed. The ICRP -84 recommend the following: i. Restricting dose to the conceptus does not mean that it is necessary for pregnant women to avoid work with radiation or radioactive materials completely, or that they must be prevented from entering or working in designated radiation areas. It does, however, imply that the employer should carefully review the exposure conditions of pregnant women. In particular, their working conditions should be such that the probability of high accidental doses and radionuclide intakes is insignificant. ii. When a medical radiation worker is known to be pregnant, there are three options that are often considered in medical radiation facilities: a) no change in assigned working duties; b) change to another area where the radiation exposure may be lower; or c) change to a job that 224 has essentially no radiation exposure. There is no single correct answer

Personnel Protection for all situations, and in certain countries there may even be specific regulations. It is desirable to have a discussion with the employee. The worker should be informed of the potential risks, local policies, and recommended dose limits. iii. Change to a position where there is no radiation exposure is sometimes requested by pregnant workers who realize that risks may be small but do not wish to accept any increased risk. The employer may also arrange for this in order to avoid future difficulties in case the employee delivers a child with a spontaneous congenital abnormality (which occurs at a rate of about 3 in every 100 births). This approach is not required on a radiation protection basis, and it obviously depends on the facility being sufficiently large and flexibility to easily fill the vacated position. iv. Changing to a position that may have lower ambient exposure is also a possibility. In diagnostic radiology, this may involve transferring a technician from fluoroscopy to CT scanning or some other area where there is less scattered radiation to workers. In nuclear medicine departments, a pregnant technician can be restricted from spending a lot of time in the radiopharmacy or working with radioiodine solutions. In radiotherapy with sealed sources, pregnant technicians or nurses might not participate in manual brachytherapy. v. An ethical consideration is involved in both of these last two alternatives since another worker will have to incur additional radiation exposure because a co-worker became pregnant. vi. There are many situations in which the worker wishes to continue doing the same job, or the employer may depend on her to continue in the same job in order to maintain the level of patient care that the work unit is customarily able to provide. From a radiation protection point of view, this is perfectly acceptable providing the foetal dose can be reasonably accurately estimated and falls within the recommended limit of 1 mGy fetal dose after the pregnancy is declared. It would be reasonable to evaluate the work environment in order to provide assurance that high-dose accidents are unlikely. vii. The recommended dose limit applies to the fetal dose and it is not directly comparable to the dose measured on a personal dosimeter. A personal dosimeter worn by diagnostic radiology workers may overestimate fetal dose by about a factor of 10 or more. If the dosimeter has been worn outside a lead apron, the measured dose is likely to be about 100 times higher than the fetal dose. Workers in nuclear medicine and radiation therapy usually do not wear lead aprons and are exposed to higher photon energies. In spite of this, fetal doses are not likely to exceed 25% of the personal dosimeter measurement. viii. Finally, factors other than radiation exposure should be considered in evaluating pregnant workers activities. In a medical setting there are 225

Textbook of Radiological Safety often requirements for lifting patients and for stooping or bending below knee level. There are a number of national groups that have established non-radiation related guidelines for such activities at various stages of pregnancy. ix. Occasionally, there are situations where family members provide essential medical care, either in the hospital or at home, to patients who have received radionuclides. In such circumstances, public dose limits do not apply to the family member. Efforts should optimally be directed at not involving females who are or may potentially be pregnant. If it is essential to involve the help of a pregnant female, it should be done in such a way that the foetal dose from this involvement does not exceed 1 mGy. Chance of Approaching the Dose Limits of Exposure Radiation doses to occupationally exposed staff working with radiological equipment are generally low and it is unlikely that the equivalent dose limit recommended by the ICRP will be approached. However, for some fluoroscopic examinations there is a potential for higher radiation doses to staff. During interventional radiology procedures, particular radiation protection problems arise from the extended fluoroscopy times and from the use of certain radiological equipments, which may not have lead rubber protective curtains. Consequently, the implications of the ICRP recommendations on the radiation exposure of the fetus of staff performing fluoroscopy procedures should be assessed. Counseling of Patients Patients who have received diagnostic studies while pregnant are often alarmed because of emotional perceptions surrounding radiation. The health professionals should advise patients about the steps that will be taken for risk assessment and provide appropriate information regarding the risk associated with diagnostic (and therapeutic) radiation exposure during pregnancy. The following points should be considered: i. It is unlikely that radiation from diagnostic radiological examinations will result in any deleterious effects on the child, but the possibility of a radiation-induced effect cannot be entirely ruled out. ii. The patient should be counseled that the risk assessment is being done not because there is reason to believe there is great risk in her circumstance but because it is one of the precautions normally taken whenever a pregnant woman receives certain diagnostic studies (Note: this applies only to diagnostic studies. The risk from therapeutic studies may be severe, such as fetal thyroid ablation). iii. Each case must be assessed according to the gestational age when exposed and the radiation levels received by the conceptus from each 226 exposure.

Personnel Protection iv. A precise fetal dose assessment requires numerous pieces of information about the X-ray system, the examinations conducted, the patient size, etc. Therefore, ‘typical’ fetal dose numbers should be used with the understanding that there may be a significant difference between the ‘typical’ dose numbers and the dose numbers resulting from an actual dose assessment. v. The dose evaluation may take up to a week to complete. vi. When all the information is acquired, the radiation risk will be assessed and will be reviewed along with other possible risks of pregnancy so that the physician, the patient, and other involved persons understand the circumstances and can thus make a reasonable decision regarding the management of the pregnancy. Other Factors Identification of the patient, female patient of reproductive age, determination of pregnancy status are necessary before performing any kind of imaging. Elimination of screening X-ray exams can significantly reduce population dose. Yearly dental check up with X-ray examinations should be avoided. Use of high speed films or DR systems in dental X-ray imaging can reduce the patient dose. Mammography examination should not be used as a screening procedure in patients with age less than 35-40, unless there is a familial history of breast cancer. The repeat X-ray exams ranges from 1 – 15%. This number will be higher in (i) training centers, due to lack of experience, (ii) mobile X-rays(chest, lumbar spine, thoracic spine, kidneys, ureter, bladder and abdomens), due to proper positioning difficulty. Lack of automatic exposure control may also increase the repeat X-rays and technique chart of various examinations should be posted at the control panel. This will enable the technologist to select correct radiographic technique. The retakes are monitored periodically and suitable action must be taken to improve image quality. Improperly loaded cassettes, excessive fog due to light leak or poor film storage conditions, processor artifacts due to dirty components or contaminated chemicals, uncalibrated X-ray unit or improper imaging techniques can also increase the retakes. A periodic quality assurance program to test the performance of the X-ray unit, image receptors and film processing systems is necessary.

NUCLEAR MEDICINE PROTECTION IN NUCLEAR IMAGING Protective Devices The radiation exposure rate in nuclear medicine ranges from 100 R/hr to 227 natural background. Hence, protective devices with suitable shielding

Textbook of Radiological Safety material should be used for personnel protection. Usually tungsten or lead are used as shielding materials in nuclear medicine. These materials are used to design the following protective devices: i. Syringe shield (Exposure decreases 100 fold) ii. Leaded glass (Radio pharmaceutical preparation areas) iii. Lead pigs (dispensing system) iv. Waste storage: Lead dust bin. The radiation levels from a medium size unshielded technetium generator having 500 mCi (18.5 GBq) would be 6 mSv / h at 0.5 m. A typical shielding used is 9.5 cm of lead, that includes the core shielding of 5.0 cm lead + additional shielding of 4.5 cm. This shielding is in higher than the minimum required shielding. For example, 6.3 cm lead shielding (9 HVL) is the minimum required to keep the radiation level at 15.0 μSv/h, in the case of Mo (HVL =7 mm lead) Syringe shielding is made up of 3 mm thick lead to reduce finger and body doses by about 200 times from syringe activities. The dose rate from an unshielded vial containing 100 mCi (4 GBq) of 99mTc would be 800 μSv/ h at 30 cm. The required minimum thickness of shielding is 1.5 mm Pb (HVL =0.25 mm Pb), to keep the surface dose level with in regulatory limits. Personnel Wear The personnel should wear laboratory coat, disposable gloves, finger ring TLD dosimeter, body dosimeters and lead apron. But lead apron has limited value in nuclear medicine as it do not attenuate medium energy photons (140 keV). Radiation Dose from Patients The nuclear medicine patients contribute radiation exposure to staff and hence reasonable distance should be kept by the staff (Table 8.8.). Patient who have received radionuclide injection should be kept in a separate areas in the department. Suitable precautions are required to reduce staff radiation dose. For example, a patient with 800 MBq of Tc-99m activity yield a exposure of 160 μSv / h at 30 cm, and 6 μSv / h at 1m respectively. Wearing a lead apron of 0.25 mm Pb may reduce the exposure from 160 to 10 μSv / h at 30 cm. Patients should be given instructions leaflet detailing contact with other people and appropriate action if clothing becomes contaminated. Table 8.8: Occupational staff exposure from patients Radionuclide

Half life

Activity (MBq)

Exposure(μSv / h at 1m)

Ga In 99m Tc 131 I 201 Tl

78 h 2.8 d 6.0 h 8.0 d 73 h

150 80 800 40 80

1.6 2.4 6.0 0.9 0.3

67

111

228

Personnel Protection Patient Dose Reduction When using radionuclides for patient studies the following points will reduce dose to the staff: i. Good counting statistics in laboratory tests ii. A diagnostic image in a reasonable time iii. Acceptable radiation dose to the patient from the target organ and excretion path way iv. Cost of expensive isotopes (123I,111I). However, the patient radiation doses in nuclear medicine studies are much lower than the radiology investigations using X-rays. For example, a urogram using iodine contrast may yield an ovarian dose of 30 mSv, bladder wall dose of 43 mSv and whole body dose of 30 mSv respectively. Pediatric Exposure The injected activity depends upon the weight and age of the patient. Guidelines are given for a standard man on the radiopharmaceutical package. Several multiplication factors are used for obtaining children doses from the adult dose as follows: i. Body surface area (BSA) ÷1.73 ii. Child’s age +1, divided by age +7 iii. Child’s weight ÷ 70 kg iv. Child’s height ÷ 174 cm. For static imaging studies first three steps (i, ii and iii) are used and for dynamic study last step (iv) is used. Most radionuclides used in nuclear medicine investigations are concentrated in breast milk. A neonate thyroid gland can receive a high radiation dose from these nuclides in mothers milk. Hence either nuclear medicine investigations should be avoided or the mother is instructed to bottle feed after the investigation for a suitable period. Over 90 % of the Tc-99m activity appears in the breast milk over 24 h and breast feeding can continue after this term. The trauma imposed on the child and mother restricting contact should be weighed against the radiation risk. Contamination Control Contamination control measures are designed to prevent radioactive material from coming into contact with personnel and prevent its spread to other work surfaces. Protective clothing (disposable plastic gloves, lab coats, closed toed shoes) and handling precautions are the basic methods of contamination control. The personnel and work surfaces are routinely surveyed for contamination and areas should be classified as radioactive and non radioactive. The work surfaces where unsealed radioactive material is handled should be covered with plastic backed

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Textbook of Radiological Safety absorbent paper. When it is contaminated or worn, it should be changed. Volatile radionuclides (e.g. I-131, Xe-133 gas) should be stored in a 100% exhaust fume hood to prevent airborne contamination and subsequent inhalation. Collimators of scintillation cameras should be covered with plastic to avoid contamination. Personnel should discard gloves in the radioactive waste dustbin after work and monitor their hands, shoes and clothing for contamination at periodic intervals. All personnel should wash their hands after handling radioactive material (before eating or drinking), to minimize internal contamination. For skin contamination, wash with soap and warm water and any aggressive decontamination may create aberrations, that can enhance internal absorption. External contamination is not a health hazard, where as internal contamination can give significant radiation exposures. To monitor the contamination control, a contamination monitor with GM type survey meter is used at the end of the day. Swipe tests (filter paper, alcohol wipes, or cotton tipped swipes) are taken on weekly basis at various locations of the nuclear medicine department. These swipe samples are counted in a NaI (Tl) gamma well counter. Areas that are in twice the background level are said to be contaminated. The contaminated areas are decontaminated followed by additional swipe tests. The GM survey is repeated to confirm decontamination. In the areas of radioactive waste and storage, the radiation levels are higher and hence exposure rate meters (ion chamber) survey is carried out to detect high exposure rates. Safety Work Practices that Can Reduce Internal Radiation Dose i. Label all radioactive containers with radionuclide name, calibration date, activity and chemical form. ii. Personnel should wear laboratory coats and gloves when handling radioactive sources. Gloves should be handled so as to avoid contamination of their inside surfaces. Lab coats, aprons, and other protective clothing should not be taken home. iii. Handbags, handkerchives, key chains etc. should not be brought inside the laboratory. iv. No eating, drinking, smoking or applying of cosmetics should occur in areas where open sources may be present. v. No foodstuffs or drinks should be stored where radioactive sources are kept, such as laboratory refrigerators. vi. Do not pipette radioactive materials by mouth. vii. Persons with an open wound should not work with radioisotope.

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Personnel Protection viii. Personnel should wash their hands after working with radioactive sources, and they should be checked for contamination by a monitor.Hands should also be monitored before going to lunch or on breaks and before leaving at the end of the day. ix. Work should be performed on absorbent pads to catch spills and prevent spattering of liquids. x. Pipettes and stirring rods should be placed in non porous trays or pans. xi. Work with radioactive gases or other volatile materials should be performed in a ventilated fume hood. These materials also should be stored in a hood. xii. Work areas should be kept tidy. Radioactive trash, contaminated pads, and so forth should be disposed of promptly. xiii. Radioactive storage areas (hot labs) should not be used to store other materials, such as office supplies or linens. xiv. Needless contamination of light switches, doorknobs, and other items that could result in unsuspected contamination to personnel should be avoided. xv. Containers with sharp or broken edges should not be used for radioactive materials. xvi. Radioactive materials should be stored when they are not in use. xvii. Discard all radioactive materials in the radioactive waste dustbin xviii. Ensure that X-133 ventilation studies are performed in a room with negative pressure with respect to hall way (if the exhaust rate is higher than the supply rate, then air will flow from hall way to room). xix. Spills or accidents should be reported to the radiological safety officer (RSO). Labeling and Identity The vial radiation shield containing radiopharmaceutical vial is labeled with radiopharmaceutical name, and patient’s name. Syringe containing a radiopharmaceutical must be labeled with radiopharmaceutical name and patient name. The patient’s identity must be verified by two ways (Name and social security number). In the case of women patient, whether she is pregnant or not is to be ascertained, by a pregnancy test. If she is a mother, breast feeding is ruled out. Activity >30 μCi, of I-131 therapeutic require a written direcetive consisting the patient identity, radionuclide, radiopharmaceutical, activity and the rute of administration. Women, nursing the infants should be advised to discondinue breast feeding, until the radioactivity in the breast milk reduces to a safe level. The recommended cessation of breast feeding periods are given the Table 8.9 and 8.10 for Tc-99m and I-31 respectively.

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Textbook of Radiological Safety Table 8.9: Cessation of breastfeeding after administration of TC-99m, radiopharmaceutical to mothers Activity

Imaging

Breast milk, safe level, μCi /ml

Cessation of breast feeding

10 mCi 5-25 mCi, Tc-kits 3-5 mCi, MAA 10-15 mCi,DTPA 5 mCi 15-25 mCi,MDP

Thyroid scan All Lung perfusion Renal scan Liver spleen scan Bone scan

8.2 x 10-2 8.2 x 10-2 1.2 x 10-1 1.2 x 10-1 1.6 x 10-1 2.1 x 10 -1

24 hr 24 hr 10 hr 17 hr 15 hr 17 hr

Table 8.10: Cessation of breast feeding after administration of I-131, radiopharmaceutical to mothers Activity

Imaging

Breast milk, safe level, μCi/ml

Cessation of breast feeding

5 μCi 10 mCi 33 mCi 100 mCi

Thyroid uptake Thyroid cancer scan Out patient therapy Thyroid cancer Treatment (ablation)

4.1 x 10-7 4.1 x 10-7 4.1 x 10-7 4.1 x 10-7

68 days Discontinue Discontinue Discontinue

PROTECTION IN RADIONUCLIDE THERAPY The treatment of thyroid cancer, and hyperthyroidism with I-131(8 days half life), require patient isolation. Once it is administered, I-131 is excreted through urine, saliva and perspiration. Hence before administration, plastic backed absorbent paper is used to cover floor, bed, mattress, light switches, toilet, and telephone. Patients meals are served by disposable meal trays. Waste containers are placed in patients room to dispose meal trays and contaminated linens. The radiation survey is made around the bedside, doorway and the neighboring room and the levels are posted for information along with instructions to nurses and visitors. The nursing staff should wear dosimeters, disposable shoe covers and gloves. The visitors are instructed to wear disposable shoe covers and the visiting times are limited Patient with I-131 therapy may be discharged once the activity <33 mCi or dose rate at 1m <0.07 mSV / hr. After the discharge the room is completely surveyed and decontaminated. It has to be confirmed by swipe tests and GM survey. If the total dose equivalent is >1mSv, the patient may be discharged with instructions. This is to minimize radiation exposure and contamination to other individuals. The radiation safety instructions that may be given to the patients are as follows:

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Personnel Protection First 4 Days i. Wash cups, plates, and eating utensils immediately after use, or use disposables. ii. Do not kiss anybody. iii. Use individual towels and wash clothes. iv. Sleep in a separate bed. v. At the end of 2 days, all personnel clothing should be washed separately. vi. Contact with infants and pregnant women should be avoided. vii. Stay 3 feet away from other people. viii. Double flesh the toilet for the first 2 days, as the urine contain large portion of activity. Next 3 Days ix. Avoid sitting close to others for hours together. This is to avoid radiation exposure from patient thyroid to others. x. Avoid holding infants / children for long periods in a day. Breastfeeding xi. Breast feeding must be discontinued. Follow the physicians advise carefully. PREGNANCY AND RADIATION PROTECTION IN NUCLEAR MEDICINE As with diagnostic radiology procedures, the pregnant patient can be apprehensive after a procedure has been performed. In the case of nuclear medicine, the patient may be even more apprehensive, realizing that an administered radioactive material has been incorporated into her body, that it will be there for some time and that it potentially may cross the placenta to the fetus. As a result of this, more careful explanation to the patient and her husband or other appropriate persons may be needed to put the potential radiation risks into perspective. In contrast to diagnostic radiography examinations, fetal doses in nuclear medicine depend predominantly upon administered activity, and they are independent of the imaging equipment. Woman of Childbearing Age and Nuclear Medicine Examinations In women of childbearing age, the possibility of pregnancy and the justification for the examination should be considered. The ICRP-84 report recommended precautions to prevent or minimize irradiation of the fetus include the following: i. The patient must be carefully interviewed to assess the likelihood of pregnancy. Particular discretion is required to ascertain the possibility 233

Textbook of Radiological Safety of pregnancy in an adolescent. In order to minimize the frequency of unintentional radiation exposures of the embryo or foetus, advisory notices should be posted at several places within the nuclear medicine department, and particularly at its reception area. For example: If it is possible that you might be pregnant, notify the physician or technician before receiving any radioactive material. ii. Many patients incorrectly assume that irradiation from a nuclear medicine examination begins when the gamma camera begins imaging, and they may not mention a potential pregnancy until after the radiopharmaceutical has been administered. Therefore, before administering radiopharmaceuticals, it is necessary to consider as pregnant any woman of reproductive age presenting herself for a nuclear medicine examination at a time when a menstrual period is overdue or missed, unless there is information that precludes pregnancy (e.g. hysterectomy or tubal ligation). If the menstrual cycle is irregular, and a non-technetium or therapeutic radiopharmaceutical is being administered, a pregnancy test may be indicated before proceeding. iii. Many laboratories also ask all females to indicate if they are breastfeeding, since many radiopharmaceuticals can be transferred to a baby via breast milk. Cessation of breastfeeding for at least some period is recommended for most nuclear medicine studies. Breastfeeding is usually stopped for 3 weeks after all 131I and 125I radiopharmaceuticals except labeled hippurate, and after 22Na, 67Ga, and 201T1. It is stopped for 12 hours after iodine labeled hippurates and all 99mTc compounds except labeled red blood cells, -phosphonates, and ±DTPA, and for at least 4 hours after the latter compounds. iv. Occasionally, questions arise about the advisability of becoming pregnant after a nuclear medicine examination or treatment. The ICRP has recommended that a woman not become pregnant until the potential fetal dose from remaining radionuclides is less than 1 mGy. This is not usually a consideration except for radioiodine therapy or radiopharmaceuticals labeled with 59Fe (for metabolism studies) or 75Se (for adrenal imaging). As a result of the long physical half-lives of these radionuclides and their long residence times in the body, it is recommended that pregnancy be avoided for 6 and 12 months respectively. The special conditions related to radioiodine therapy are discussed later. Actions and Precautions that can Reduce Radiation Exposure to the Fetus Pregnancy need not be considered a contradiction for nuclear medicine procedures, in particular diagnostic ones involving short lived radionuclides, provided there are strong clinical justifications and effort 234 has been made to explore alternatives involving non-ionizing radiation.

Personnel Protection Since radionuclides in maternal tissues contribute to fetal dose, maternal hydration and frequent voiding can reduce the fetal dose after the administration of a number of radiopharmaceuticals. Irradiation of the fetus results from placental transfer and distribution of radiopharmaceuticals in the fetal tissues, as well as from external irradiation from radioactivity in the mother’s organs and tissues. The physical, chemical, and biological properties of the radiopharmaceuticals are the critical factors in possible placental transfer. Using smaller administered activities and longer imaging times can reduce the absorbed dose to the foetus. This is feasible if the patient is not too sick and is able to remain still. Occasionally, the sequence of the examinations and choice of radiopharmaceutical can be adjusted to reduce radiation dose. Typical fetal doses from nuclear medicine examinations for common radiopharmaceuticals are presented in the following Table 8.11. Table 8.11: Fetal whole body dose from common nuclear medicine examinations in early pregnancy (Dose includes maternal and fetal self-dose contributions). (Adapted from Russell, Stabin, Sparks, et al. 1997, ICRP 53, and ICRP 80) Radiopharmaceutical

Procedure

Tc Tc 99m Tc 99m Tc 99m Tc 99m Tc 99m Tc 67 Ga 123 I 131 I 131 I

Bone scan (phosphate) Lung perfusion (MAA) Lung ventilation (aerosol) Thyroid scan (pertechnetate) Red blood cell Liver colloid Renal DTPA Abscess / tumor Thyroid uptake1) Thyroid uptake1) Metastases imaging1)

99m 99m

1)

Administered activity (MBq)

Early pregnancy (mGy)

Nine months (mGy)

750 200 40 400 930 300 750 190 30 0.55 40

4.6-4.7 0.4-0.6 0.1-0.3 3.2-4.4 3.6-6.0 0.5-0.6 5.9-9.0 14-18 0.4-0.6 0.03-0.04 2.0-2.9

1.8 0.8 0.1 3.7 2.5 1.1 3.5 25 0.3 0.15 11.0

Foetal thyroid doses are much higher than foetal whole body dose, viz. 5-15 mGy / MBq for I and 0.5-1.1 Gy / MBq for 131 I.

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RADIOIODINE THERAPY AND PREGNANCY Treatment of Pregnant Patients with Radionuclides As a rule, a pregnant woman should not be treated with a radioactive substance unless the radionuclide therapy is required to save her life: in that extremely rare event, the potential absorbed dose and risk to the foetus should be estimated and conveyed to the patient and the referring physician. Considerations may include terminating the pregnancy. Radioiodine therapy is essentially contraindicated in patients who are known to be pregnant. Radioiodine easily crosses the placenta and the fetal 235

Textbook of Radiological Safety thyroid begins to accumulate iodine at about 10 weeks of gestational age. If radioiodine treatment of thyroid carcinoma is to be performed, it should be delayed until after delivery. If this is done, the physician should also be aware that radioiodine is excreted in breast milk and breastfeeding should be stopped completely after a therapeutic dose. If this is not done the infant may become permanently hypothyroid or be at high risk for subsequent thyroid cancer. In women, thyroid carcinoma comprises over 80% of cancer of the head and neck diagnosed between the ages of 15 ± 45 years. Thyroid cancers are relatively non-aggressive compared to most other cancers. As a result both surgical and radioiodine treatment are often delayed until after pregnancy. In general, if any therapy is to be performed during pregnancy, it will be surgery during the second or third trimester. Steps to be taken in Patients Found to be Pregnant after Administration of Radioiodine Therapy A major problem occurs when a female, who is not thought to be pregnant, is treated for thyroid carcinoma and is found to be pregnant after the administration of radioiodine. Menstrual history is often not adequate to ensure that a patient is not pregnant. In most developed countries, it is common practice to obtain a pregnancy test prior to high-dose 131I scanning or therapy for women of childbearing age unless there is a clear history of prior tubal ligation or hysterectomy precluding pregnancy. In spite of the above, it still happens that pregnant women are treated, either because of false histories or because the pregnancy is at such an early stage that the pregnancy test is not yet positive. Most commonly, the pregnancy is early and the major problem is fetal whole body dose due to gamma emissions from radioiodine in the maternal bladder. During pregnancy, the whole body dose to the conceptus is in the range of 50±100 mGy/MBq of administered activity. This dose can be reduced by hydrating the patient and by encouraging frequent voiding. If the conceptus is more than 8 weeks post conception (and the foetal thyroid may accumulate iodine) and the pregnancy is discovered within 12 hours of iodine administration, giving the mother 60 ± 130 mg of stable potassium iodide (KI) will partially block the foetal thyroid and reduce thyroid dose. Twelve hours after radioiodine administration, this intervention is not very effective. Risk to a Pregnant Woman, When a Family Member is Treated with Radioiodine Patients treated with radioiodine can be a significant radiation source to pregnant family members. The dose to a family member staying at a distance 236 of 0.5 meters from a patient until the radioactivity totally decays (about 10

Personnel Protection weeks) is about 1.3 mGy from a hyperthyroid patient and 6.8 mGy from a thyroid cancer patient. Also, these patients must be careful not to transfer radioiodine contamination to family members by direct or indirect means. Avoid of Pregnancy after Radionuclide Therapy Most female patients are advised not to become pregnant for at least 6 months after radiotherapy with radioiodine. This is not based upon potential heritable radiation effects, but rather upon the need to be sure (1) that the hyperthyroidism or cancer is controlled, and (2) that another treatment with radioiodine is not going to be needed when the patient is pregnant. It is also based upon the fact the ICRP has recommended that enough radioiodine be cleared to ensure that the unborn child does not receive a dose in excess of 1 mGy unless it is medically necessary for the health of the mother. There are occasional circumstances in which 32P, 89Sr, or 131I metaiodobenzylguanidine are used for therapy. In order to keep the dose to the foetus below 1 mGy, pregnancy should be avoided for 3, 24, and 3 months respectively. STAFF PROTECTION Pregnant Staff and Continuation of Work For most diagnostic procedures, there would be no need for pregnant staff to take any additional precautions other than limiting their direct contact to as short as necessary. Because the exposure from patients who have been administered radiopharmaceuticals is quite low, there is no radiological reason not to continue imaging procedures. Moreover, the International Basic Safety Standards require that “notification of pregnancy shall not be considered a reason to exclude a female worker from work”. If the institution administers therapeutic activities of 131I for thyroid cancer, should be considered refraining from this particular part of the practice. Not only that there are higher exposures associated with the administration of 131I, but also iodide is volatile, increasing the potential for internal uptake. Some radiopharmaceuticals cross the placenta freely, e.g. radioactive iodides, and are taken up in foetal tissues, where they irradiate the tissues. Some analogues of natural metabolites (e.g. radiostrontium for calcium and radiocaesium for potassium) are less readily transferred. Radiopharmaceuticals that are retained by the mother, and do not cross the placenta (e.g. radiocolloids), only act as external sources of irradiation to the fetus. Maternal Hydration In the case of radiopharmaceuticals that are rapidly eliminated by the maternal kidneys, the urinary bladder, acting as a reservoir, is a major source of foetal irradiation. After the administration of such radiopharmaceuticals, 237 maternal hydration and frequent voiding should, therefore, be encouraged.

Textbook of Radiological Safety For those radiopharmaceuticals that have gastrointestinal excretion however, administration of laxatives is only rarely helpful in reducing fetal dose. A typical example of sequence is a ventilation perfusion lung scan ordered on a pregnant patient to exclude a pulmonary embolus. In routine operation, many laboratories will perform the ventilation scan first and then do the perfusion scan. This has advantages in some situations. In the specific case of a suspected pulmonary embolus, the perfusion scan can be performed first, and if the result is normal, a ventilation scan is not needed at all. The choice of radiopharmaceuticals for the ventilation portion of the lung scan can also affect fetal dose. If the scan is performed with 133Xe gas, there is very little fetal dose; however, one can also do ventilation scans using 99mTc-DTPA aerosol. This will be absorbed and excreted via the kidneys, and while in the bladder it will contribute to fetal dose.

RADIOTHERAPY Teletherapy equipment or Brachytherapy equipment, including sources and accessories, shall not be used unless all the relevant quality assurance tests have been satisfactorily performed. These tests should be repeated periodically and their records are maintained in the form of log book. Appropriate equipments and accessories must be made available for performing dosimetry, treatment planning, simulation and quality assurance. These instruments are calibrated and maintained in good working condition. In Brachytherapy the applicator / sources in the patient must be verified with the help of radiography or autoradiography. The position of the sources are retained in the same way through out the treatment time. If the treatment time is in the order of minutes, then the operator should watch the treatment, in order to terminate the treatment with in the intended time. PERSONNEL SAFETY DURING SOURCE TRANSFER OPERATIONS OF TELETHERAPY AND HDR BRACHYTHERAPY UNITS Owing to the decay of the radiation sources, in radiotherapy units, replacement of the decayed sources with fresh sources of higher activity is necessary and the frequency of replacement depends upon the half life of radioactive source used (e.g. 5 yrs. for cobalt-60). Sometimes, the repair work of the unit’s head also demands source transfer operation. The activity of these sources is dependent on the make and model of the therapy unit. The activity of a teletherapy source is much higher that that in a remotely controlled after loading brachytherapy unit. Therefore, more attention is needed during the source transfer of the teletherapy units. Ignoring any of 238 the recommended procedures and steps may result in a serious radiation accident and elaborate and expensive management of the situation.

Personnel Protection Make and Models Different types and makes of teletherapy and brachytherapy units are installed and subsequently serviced by various manufacturers, all over India. Table 8.12 gives details about some of the teletherapy units available in the country and the type of source drawer of each mode, which is replaced during the source transfer operation. Table 8.13 gives the details about the remotely controlled brachytherapy units available in India. Table 8.12: Teletherapy units Model of the unit

Type of source drawer

Local agents

Theratron-780 780C, 780E, Phoenix Theratron-1000E Equinox, 80/100 Shinva FCC 8000 Shinva FCC 8000 F Bhabhatron-II

Cylindrical

M/s Kirloskar Theratronics Pvt, Ltd., New Delhi

Cylindrical

M/s. Indelelect Technologies Chennai M/s Panacea Medical Technologies Pvt.Ltd., Bengaluru UB Electronics Instruments Ltd. Varian Medical India Pvt. Ltd., Mumbai

Cylindrical

ATC-C9

Capsule

Alcyon

Head @

@ : Source is loaded in unit’s head.

The new sources loaded in drawers or capsules are transported in special source flasks designed for the particular make and model of the therapy unit. The source flask is used to carry out the source transfer operation. The decayed source is first removed from the unit and loaded into source flask. New source is then loaded into the unit. During repair works (involving head of the unit) also, the source flask is used for removing the source from the unit and for reloading it into the therapy unit on completion of the repair work. Table 8.13: Brachytherapy units Model of the Unit Selectron LDR Microselectron HDR Microselectron HDR Genie Gamma Med plus iX Varisource iX Flexitron Gynie and Multisource

Local agents M/s. Nucletron India Pvt. Ltd.,Chennai

M/s. Varian Medical Systems India Pvt. Ltd. Mumbai M/s.Isodose Control,Tiruchirapalli, BEBIG, Eckert and Ziegler,Chennai

239

Textbook of Radiological Safety General Precautions 1. It is the responsibility of serving company to ensure safe transport of the source flask loaded with new source either supplied by Board of Radiation and Isotope Technology (BRIT), Mumbai or imported and the return of the decayed source back to the supplier. All the regulations for safe transport should be strictly followed. 2. All the gadgets required for the source transfer operation as per the check list should be made available before undertaking the operation. Nonstandard gadgets which may delay the source transfer operation and may lead to radiation accidents should not be used. 3. Personnel monitoring badges, Pocket dosimeters and Pocket alarm type radiation monitor should be used by all personnel involved in the operation. 4. Proper procedure should be adopted for the alignment of the source flask with the unit head. Necessary coupling devices should be used to facilitate proper alignment. 5. Proper arrester/s should be provided on the source flask after removing the lids so that source drawer does not slip down accidentally during the alignment process. 6. After confirming the alignment between the source flask and unit head, the source flask should be immobilized. Also, all unnecessary materials around the flask should be removed. 7. In case the source transfer operation is to be carried out in the ON position (like in Philips or Picker units), proper locking device should be provided to prevent the movement of the system in case of power failure. 8. Wherever dummy drawer is required in the source transfer operation (theratron and Siemen’s Model), source transfer operation without dummy drawer should not be carried out. 9. After the transfer of the new source into the teletherapy unit, proper arrester or arresting plates should be employed to prevent the slipping of the source drawer, in case the unit is moved accidentally. 10. Proper functioning of the unit, after loading of new source, should be checked with shutter open to minimum size and beam pointing opposite to maze wall. 11. Parking rod required for pushing the stuck source drawer or shutter should always be available in the control room. PREGNANCY AND RADIATION PROTECTION IN RADIOTHERAPY Special care needs to be taken when considering radiotherapy for patients during pregnancy. There are no hard and fast rules, but the patient, family 240 members, treating oncologist and other team members should carefully discuss the decisions made.

Personnel Protection Pregnant Woman and Radiotherapy Pregnant patients can be treated with radiotherapies, but important factors must be considered. The most important considerations, as suggested by the ICRP, include: 1. The stage and aggressiveness of the tumor; 2. The location of the tumor; 3. Potential hormonal effects of pregnancy on the tumor; 4. Various therapies and their length, efficacy, and complications; 5. Impact of delaying therapy; 6. Expected effects of maternal ill-health on the fetus; 7. The stage of pregnancy; 8. Fetal assessment and monitoring; 9. How and when the baby could be safely delivered; 10. Whether the pregnancy should be terminated; 11. Legal, ethical, and moral issues. Reduction of Fetal Dose When a Pregnant Patient Undergoes Radiotherapy The first consideration is if the treatment can be postponed until the fetus is at a later gestation age. If the decision is made that radiotherapy is necessary, it is important to calculate the dose to the fetus before the treatment is given. When external radiotherapy is used for treatment of tumors at some distance from the fetus, a very important factor in fetal dose is the distance from the edge of the radiation field. The American Association of Physicists in Medicine (AAPM) recommends that the following points for consideration: 1. Complete all planning as though the patient were not pregnant. If the foetus is near the treatment beam, do not take portal localization films with open collimation and blocks removed. 2. Consider modifications to the treatment plan that would reduce the radiation dose to the fetus (e.g. changing field size, radiation energy). If possible use photon energies of less than 25 MV. 3. Estimate dose to the fetus using phantom measurements, a shield may be constructed with 4-5 half-value layers of lead. 4. Document the treatment plan and discuss it with the staff involved in patient set-up. Document the shielding (perhaps with a photograph). 5. Check the weight- and load-bearing specifications of the treatment couch or other aspects of shielding support. 6. Be present during initial treatment to assure that shielding is correctly placed. 7. Monitor the fetal size and growth throughout the course of treatment and reassess fetal dose if necessary. 8. On completion of treatment, document the total dose including the range 241 of dose to the fetus during therapy.

Textbook of Radiological Safety 9. Consider referring the patient to another institution if equipment and personnel are not available for reducing and estimating the fetal dose. Pregnant Patient with Cervical Carcinoma Who Needs Radiotherapy Treatment The patient diagnosed as cervical cancer may be treated with radiotherapy with external and Brachytherapy treatment. If she is found to be pregnant unfortunately, it is likely that pregnancy will be terminated. Carcinoma of the cervix is the most common malignancy associated with pregnancy. Cervical cancer complicates about one out of 1250 to 2200 pregnancies. This rate, however, varies significantly by country. Cervical cancer is often treated by surgery/radiotherapy (external beam radiotherapy and brachytherapy) and the absorbed doses required with both forms of radiotherapy will cause termination of pregnancy. If the tumor is infiltrative and is diagnosed late in pregnancy, an alternative is to delay treatment until the baby can be safely delivered. Regardless of protective measures, radiotherapy involving the pelvis of a pregnant female almost always results in severe consequences for the fetus, most likely fetal death. Avoid of Pregnancy after Receiving Radiotherapy for Breast Cancer Treatment The wait can be substantial and needs to be discussed with her radiation oncologist. Most radiation oncologists advise their patients not to become pregnant for 1-2 years after completion of therapy. This is not primarily related to concerns about potential radiation effects, but rather to considerations about the risk of relapse of the tumour that would require more radiation, surgery, or chemotherapy. Illustration 1 A patient just finished four weeks of radiation treatment to the neck area for non-Hodgkin’s lymphoma. After one month, she got pregnant. What are the possible effects on the fetus? There is no likely effect. Radiation exposure occurred before conception, so any effect on the offspring would be classed as genetic effect. No data from humans show any statistically significant genetic effect in any population, even the Japanese atomic bomb survivors. All estimates of genetic radiation risk come from studies of rodents, which show that males are far more sensitive than females. However, it is not easy to extrapolate this data to humans. There is very low risk of any effect on the unborn child. The World Health Organization estimates that the worldwide incidence of inherited disease (ranging from severe to as trivial as an inconspicuous birthmark) is about 10%. In the unfortunate event that the 242 child is born with any genetic abnormality, it is extremely unlikely that it would be related to the earlier radiation exposure.

Personnel Protection Illustration 2 What is the risk to a foetus if a man who has had a radiation pellet inserted into his prostate for cancer treatment comes into close proximity with a pregnant woman? There is no danger involved. Prostate brachytherapy can be performed with permanent implantation of radioactive 103Pd or 125I seeds, and the patient is discharged from hospital with these in place. The short range of the emissions from these radionuclides is the reason that the patient can be discharged and is the reason that these patients pose no danger to pregnant family members. Other brachytherapy patients are kept in the hospital until the sources are removed. While these patients can occasionally be a source of radiation to a pregnant family member, the potential dose to the fetus is very low, irrespective of the type of brachytherapy. RECORDS Treatment records must be maintained to facilitate unambiguous and correct treatment follow up. In teletherapy the following records and log books are required. 1. Patient case sheet 2. Patient simulation register 3. Patient planning (TPS) register 4. Patient Treatment register 5. QA register (daily/quarterly/annually) 6. Calibration register 7. Radiation survey register 8. Machine log book (for each machine) In brachytherapy also all the above records and log books are required. In addition, the following log books are maintained 9. Source inventory register 10. Source movement register 11. Source disposal register The number of logbook and type of entries varies with the machine and type of treatment. Every machine should have a separate machine log book and the entries are made as follows in Table 8.14: Table 8.14: Model log book Date

defect

Reported to whom

solutions

Serviced by

Down time Instructions, Physicist If any /RSOsign

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Textbook of Radiological Safety BIBLIOGRAPHY 1. Europian commission, European Guidelines on Quality Criteria for Diagnostic Radiographic Images in Paediatrics, EUR-16261, Luxembourg 1996. 2. International commission on radiological protection, 1990 Recommendations of the ICRP, Publication 60, Pergamon Press, Oxford 1991. 3. International commission on radiological protection, Pregnancy and Medical Radiation, Annals of the ICRP, Publication 84, Pergamon Press, Oxford 2000. 4. International commission on radiological protection, Radiation Dose to Patients from Radiopharmaceuticals, ICRP Publication 53, Pergamon Press, Oxford 1988. 5. International commission on radiological protection, Radiation Dose to Patients from Radiopharmaceuticals, ICRP Publication 80, Addendum to ICRP 53, Pergamon Press, Oxford 1998. 6. International commission on radiological protection, Pregnancy and Medical Radiation, ICRP Publication 84, Pergamon Press, Oxford and New York 2000. 7. International commission on radiological protection, Managing Patient Dose in Digital Radiology, ICRP Publication 93, Pergamon Press, Oxford and New York 2004. 8. National radiological protection board, Doses to Patient from Medical X Ray Examinations in the UK: 2000 review, NRPB-W14, Chilton 2002. 1. 9. Russell JR, Stabin MG, Sparks RB. Radiation absorbed dose to the embryo/ foetus from radiopharmaceuticals, Health Phys 1997;73:756-69. 10. Stovall M, Blackwell CR, Cundiff J, Novack DH, Palta JR, Wagner LK, et al. Foetal dose from radiotherapy with photon beams: Report of AAPM Radiation Therapy Committee Task Group No. 36, Med. Phys. 1995;22(1):63-82.

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Chapter

9

Transport of Radioactive Materials

INTRODUCTION Radioactive materials are transported in a variety of types of packages. Transport includes any operation incidental to the whole course of carriage, such as loading, unloading and storage in transit. There is a need for certain safety requirements during such transports. These requirements must ensure the safety of persons, property and the environment against radiological hazards involved in transport. To minimize the hazards during transport of radioactive materials and to ensure safe transport of radioactive materials several procedures are adopted, which include; 1. Limiting the amount of radioactive material in a package, depending on the ability of the package to withstand both normal and accidental conditions encountered during transport, 2. Limiting the radiation level on the surface of the package and at a distance of 1 meter from the surface of the package, and 3. Segregating such packages from passenger areas and undeveloped photographic films. The segregation distance is such that, it should limit the exposure of undeveloped photographic film to 0.1mSv per consignment of such film. 4. The Types of package must be specified and the quantity of radioactive material in a package must not exceed the activity limits specified by the competent authority. 5. Packages must be designed to confirm the category and Transport index in accordance with the conditions and requirements, that are specified. 6. Each package of gross weight exceeding 30 kg must have its gross weight legibly and durably marked on the outside of the packaging 7. The labels must be fixed on two opposite sides of the outside of a package or on the outside of all four sides of a freight container or tank. 8. Large freight containers carrying packages other than excepted packages, and tanks must bear four placards. 9. Each consignment of radioactive material is accompanied with declaration document as specified.

Textbook of Radiological Safety 10. In case of emergency, the local police should be informed, to initiate steps to cordon off the area. The competent authority should be informed within 24 hours. Radioactive materials should be properly packed prior to transport and should be accompanied by appropriate transport certificates. Prior to undertaking the transport of radioactive material, the consigner should approach the Radiological safety division, AERB for permission and advice. The Atomic Energy Regulatory Board (AERB) has issued certain guidelines in the form of codes SC/TR-1, SC/TR-3, and SG/TR-3 for transport of radioactive materials. The main objective of the above code are that, the doses to persons engaged in transport and other persons are as low as reasonably achievable, during such transport. TYPES OF PACKAGES Radioactive packages are classified into five types depending on total activity in a package, specific activity of the material, and whether the material is fissile or not. Sturdiness, shielding integrity of the packaging and the quantity of radioactive material determine the type of the package. The package types are; a. Type A b. Type B c. Excepted packages d. Industrial packages e. Fissile packages. Type A Packages Type A packages are required to withstand the normal conditions of transport without loss or dispersal of their contents or loss of adequate shielding integrity. The total activity of radioactive material in Type A package is limited. Type A packages must not contain activities greater than A1 for special form radioactive material or A2 for all other radioactive material. A1 is the maximum activity of special form radioactive material permitted under provisions of the type A package. A2 is the maximum activity of radioactive material, other than special forms of radioactive material permitted under provisions of the Type A package. The values of A1 and A2 of important clinical radionuclides are given in the Table 9.1 (AERB safety code SC/TR-1,1986). The special form of radioactive material means either an indispersible solid radioactive material or an ordinarily unbreakable metallic sealed capsule containing the radioactive material. The sealed capsule should be so constructed that it can be opened only by destroying the capsule. Special radioactive material should (i) have at least one dimension not less than 5 246 mm, and (ii)comply with the prescribed test requirements.

Transport of Radioactive Materials Table 9.1: A1 and A2 values for clinical radionuclides Radionuclide Co-60 Cs-137 I-131 Ir-192 P-32 Ra-226 Sr-90 Tc-99m I-125

A1 (TBq)

A1 (Ci)

A2 (TBq)

A2 (Ci)

0.4 2 3 1 0.3 0.3 0.2 8 20

10 50 80 20 8 8 5 200 500

0.4 0.5 0.5 0.5 0.3 2 × 10-2 0.1 8 2

10 10 10 10 8 5 × 10-1 2 200 50

Type B Packages Type B packages are designed to withstand both normal and accidental conditions of transport. Radioactive materials in quantities greater than those allowed in Type A packages are shipped in Type B packages. The design of the packages are subject to the approval of the competent authority. Type B packages must not contain (i) activities greater than those authorized for the package design (ii) radionuclides different from those authorized for the packages design or (iii) contents in a form, or a physical or chemical state different from those authorized for the package design, as specified in their certificates of approval. Type B packages are divided into two basic categories namely Type B(U) and Type B(M).The type B(U) packages meet the most stringent requirements; it is considered that safety is entirely built in to these packages. The design of this package is subject to the approval by the competent authority of the country of origin. Type B(M) packages do not meet all the requirements applicable to Type B(U) packages; however, they must incorporate alternative design features and operational controls must be instituted so as to achieve the same level of safety as for Type B(U). The design and shipment of Type B (U) package is subject to multilateral approval (country of origin and the country through which it is transported). Excepted Packages Excepted package means a packaging containing excepted quantities of radioactive material that is designed to meet the general requirements for all packaging and packages. These packages are permitted to contain small quantities of radioactive materials and are excepted from various specific packaging and labeling requirements. An excepted package must not contain activities greater than A1 and A2 values of the radionuclides. For transport by post, the total activity in each package must not exceed one tenth of the above relevant limit specified. 247

Textbook of Radiological Safety Industrial Packages The total activity in a single package of low specific activity (LSA) material or in a single package of surface contaminated object (SCO) must be so restricted that the external radiation level at 3 m from the unfinished material or object or collection of objects does not exceed 10 mSv/h and the activity in a single package must also be so restricted that the activity limits for a conveyance specified by the competent authority. LSA material means radioactive material which by its nature has a limited specific activity, or radioactive material for which limits of estimated average specific activity apply. External shielding materials surrounding the LSA materials must not be considered in determining the estimated average specific activity. The LSA has three groups namely LSA-I, LSA-II and LSA-III. SCO means a solid object which is not itself radioactive but which has radioactive material distributed on its surfaces. SCO has two groups namely SCO-I and SCO-II. Fissile Packages Fissile material means uranium-233, uranium-235, plutonium-238, plutonium-239, plutonium-241 or any combination of these nuclides. Criticality safety in transport of fissile material is ensured by limiting the quantity and geometric configuration of the fissile material, package design features and by controlling the number of packages to be carried on a single conveyance or to be stored together in transit. All packages containing fissile material must comply with the applicable activity limits specified in the safety code. Packaging containing fissile material must not contain (i) a mass of fissile material greater than that authorized for the package design (ii) any radionuclide or fissile material different those authorized for the package design or (iii) contents in a form or physical or chemical state, or in a spatial arrangement, different from those authorized for the package design, as specified in their certificates of approval. The packages containing fissile materials, which may be LSA, SCO, Type A or Type B, are subjected to multilateral approval. TRANSPORT INDEX The Transport index (TI) of a package is the number expressing the maximum radiation level in mrem /h at 1m from the external surface of the package. If the radiation level is measured in mSv /h, it should be multiplied by 100,to arrive mrem /h. The values of TI are 0, 1-10 and more than 10.

248

Transport of Radioactive Materials Categories There are three categories of packages namely (i) category I-White, (ii) category II-yellow, and (iii) category III-Yellow. Category I-white The maximum radiation level at any point on the external surface of the package should not be more than 0.005 mSv/h or 0.5 mrem/h. The Transport index is zero. Category II-Yellow The maximum radiation level at any point on the external surface of the package should be more than 0.005 mSv/h, but not more than 0.5 mSv/h or 50 mrem/h. The Transport index is more than zero, but not more than 1. Category III-Yellow The maximum radiation level at any point on the external surface of the package should not be more than 0.5 mSv/h, but not more than 2 mSv or 200 mrem/h. The Transport index is more than 1, but not more than 10. PACKAGING AND PACKAGE REQUIREMENTS The package should satisfy the following general and specific requirements before they are used for actual transport. General Requirements 1. The packages must be so designed in relation to its mass, volume and shape that, it could be easily and safely handled and transported. 2. The lifting attachments on the package should not fail when used in the indented manner. 3. The external surface of the package should be free from protruding features and can easily decontaminated. 4. The outer layer of the package must be designed as to prevent the collection and the retention of water. 5. The package should be capable of withstanding the effects of acceleration, vibration etc. during transport. 6. The material of the packaging and the components must be physically and chemically compatible with one another, and with radioactive contents. 7. All valves through which the radioactive contents could otherwise escape must be protected against unauthorized operation. 8. Packages to be transported by air, should with stand ambient temperatures ranging from – 40°C to + 55°C 9. Packages with liquid radioactive material, transported by air, must with 249 stand an internal pressure of not less than 95 kPa.

Textbook of Radiological Safety Additional Requirements for Type A Packages 1. The smallest overall external dimension of the packaging must not be less than 10 cm. 2. The outside of the package must incorporate a seal, to serve as proof against tamper. It will prevent unintentional release of radioactivity. 3. Any tie-down facility on the package should withstand both normal and accidental conditions. 4. The design of the package must take into account temperatures ranging from 40-75°C, for the components of the packaging. 5. The design, fabrication and manufacturing techniques must be in accordance with national and international standards. 6. The containment system is securely closed by a positive fastening device. It can not be opened unintentionally. 7. The package containment system must retain its radioactive contents under a reduction of ambient pressure to 25 kPa. 8. When the packages are subjected to specified tests, it would prevent (i)loss or dispersal of the radioactive material, and (ii) loss of shielding integrity, which may result >20% radiation level on the external surface of the package. 9. If the packages carries liquid radioactive material, it should accommodate variations in the temperature of the contents, dynamic effects and filling dynamics. 10. If the liquid volume is less than 50 ml, it is provided with sufficient absorbent material to absorb twice the volume of the liquid contents. Test for Type A Package The Type A package tests are (i) water spray test, (ii) free drop test, (iii) stacking test, and (iv) penetration test. These tests will demonstrate the ability of the package to withstand normal conditions of transport. After the specimen is subjected to tests, suitable assessment must be made to assure that the requirements of the packages are fulfilled in conformance with the performance and acceptance standards. Water Spray Test The specimen must be subjected to a water test that simulate exposure to rainfall of approximately 5 cm per hour for at least one hour. Free Drop Test

250

The specimen must drop on to the target so as to suffer maximum damage in respect of the safety features to be tested. The height of drop measured from the lowest point of the specimen to the upper surface of the target must be not less than 1.2 m for packages weighing up to 5000 kg.

Transport of Radioactive Materials Stacking Test The specimen must be subjected to a compressive load equivalent of 5 times the mass of the actual package for a period of 24 hours. Penetration Test The specimen is placed on a rigid flat horizontal surface. A bar of 3.2 cm diameter with a hemispherical end and a mass of 6 kg must be dropped and directed to fall, with its horizontal axis vertical, on to the centre of the weakest part of the specimen. The height of drop of the bar must be 1m. Additional Requirements for Type B Packages 1. The package should meet the requirements of general requirements 1-8 and that of Type A package (1-10). 2. The packaging is so designed, that if it were subjected to the prescribed tests it would retain sufficient shielding to ensure that the radiation level at 1 m from the surface of the package would not exceed 10 mSv/h with the maximum radioactive contents which the package is designed to carry. 3. The package must withstand heat generated within the package, so that the heat should not (i) alter the arrangement, geometrical form or the physical state of the radioactive contents, (ii) lessen the efficiency of the packing, and (iii) accelerate corrosion in combination with moisture. 4. If it were subjected to water spray test, free drop test, compression test and penetration test the loss of radioactive contents should be less than A2 × 10-6 per hour. 5. If the package is subjected to mechanical test, thermal test and water immersion test, it would restrict the accumulated loss of radioactive contents in a period of one week to not more than 10 × A2 for Krypton-85 and not more than A2 for all other radionuclides. Test for Type B Package The test for Type B packages are (i) mechanical test, (ii) thermal test and (iii) water immersion test. These tests will demonstrate the ability of the package to withstand accident conditions in transport. After the specimen is subjected to tests, suitable assessment must be made to assure that the requirements of the packages are fulfilled in conformance with the performance and acceptance standards. Mechanical Test This consists of three different drop tests. For drop I, the specimen must be dropped from 9 m onto the target so as to suffer the maximum damage. For drop II, the specimen must be dropped from 1 m, so as to suffer the 251 maximum damage onto a bar rigidly mounted perpendicularly on the

Textbook of Radiological Safety target. The bar must be a solid mild steel of circular section (15 ± 0.5 cm) in diameter and 20 cm long. For drop III, a 500 kg mass is dropped from 9 m height onto the specimen. The mass must consist of a solid mild steel plate 1m by 1m and must fall in a horizontal altitude. Thermal Test The specimen must be subjected to a fuel source (hydrocarbon /air fire), having a emissivity coefficient of 0.9 and average flame temperature of 800°C for a period of 30 minutes. The fuel source must extend horizontally at least 1m, and must not extend more than 3 m, beyond the external surface of the specimen. The specimen must be positioned 1m above the surface of the fuel source. After the thermal test, the specimen must not be cooled artificially and any combustion materials of the specimen must be allowed to proceed naturally. Water Immersion Test The specimen must be immersed under a head of water of at least 15 m for a period of not less than 8 hours in the altitude which will lead to a maximum damage. For demonstration purposes, an external gauge pressure of at least 150 kPa must be considered to meet these conditions. Package Handling Packages weighing <5 kg are not provided with handling facilities, and one can hold them with hands. Packages weighing not more than 30 kg, is provided with manual handling facility. Two or more persons should be employed for handling the packages up to 100 kg. During manual handling, operations should be completed quickly, by keeping the packages as far away as possible. Mechanical handling devices (crane, chain pulley block etc.) should be employed to handle the packages, if the weight is > 100 kg. If the weight of the package is not known, it must be handled only by mechanical means. Storage in Transit Requirements 1. Packages of radioactive materials should be kept segregated from areas routinely occupied by passengers and public. 2. The number of packages stored in an area should be restricted to ensure that the sum of the transport indexes of the packages stored in the area does not exceed 50. 3. Radioactive consignments should not be stored together with other dangerous goods (explosives, inflammables etc). 4. They should also be so separated from underdeveloped X-ray and photographic films or plates so that these are not exposed to more than 0.1mGy (10 mR), during the entire duration of the transport including 252 storage in transit.

Transport of Radioactive Materials 5. Persons working in the storage area should spend minimum possible time in the vicinity of the packages, relevant to their work. 6. When the consignment is brought for booking, it should be forwarded for transport as soon as possible. 7. The packages must be delivered to the consignee or his authorized representative only. 8. An unclaimed or damaged package should not be auctioned or otherwise disposed off. 9. If the package is damaged, do not touch the package, isolate it by cordoning about 5 m around it, and inform the competent authority. Package Radiation Levels Radiation levels under normal transport conditions are limited so that maximum radiation level at the package surface should not be greater than 2 mSv/h (200 mR/h) and the maximum radiation level at 1m from the surface should not be greater than 0.1 mSv/h (10mR/h). PREPARATION OF THE PACKAGE FOR TRANSPORT 1. The source should be transported only in an approved transport container. Alternatively the original container in which the fresh source was received can be used, provided if it is in good condition. 2. The source should be loaded in the container properly and carefully. 3. The source should be secured within the shielded container by means of appropriate locking mechanisms incorporated in the design of the shielded container. 4. The lid of the container should be closed, so that the source is not released during the transport. 5. The container should be loaded in an outer container such as wooden or metallic box. It is provided with spacers within for preventing movement of the shielded container inside during transport. It should be ensured that the outer container deployed is in a good condition and is provided with locking facility and strong lifting handles. 6. The outer container should be locked and tied with crossed metal straps and sealed, then only it is called transport package. 7. The maximum radiation levels on the outer surface of the package and at a distance of 1m from the surface should be measured (mSv/h) by using a working radiation survey meter and recorded. This will be useful to find the correct transport index (TI). The measured value is multiplied by 100, to get the appropriate TI. MARKING OF THE PACKAGE Write or inscribe the following information durably, clearly and legibly on 253 the outer side of the package.

Textbook of Radiological Safety 1. Addresses of the CONSIGNOR and the CONSIGNEE. 2. Type of package (e.g. Type A/Type B etc.). 3. UNITED NATIONS NUMBER (UN NO.), and the PROPPER SHIPPING NAME (please refer Table 11.3). 4. Gross weight of the package if it exceeds 30 kg for domestic transport and 50 kg for international transport. 5. Competant authority (i) identification mark allocated to that design and (ii) serial number to identify each package, if it is a Type B(U)/B(M) package. 6. In the case of Type B(U) or B(M), the outer surface should have the trefoil symbol (Fig. 9.1), which must be marked by embossing, stamping or other means resistant to the effects of fire and water.

Fig. 9.1: The trefoil symbol

LABELING OF THE PACKAGE Appropriate labels indicating the category of the package should be affixed on two opposite sides on the exterior of each package. In the case of a tank or freight container it should be on the outside of all four sides. Each label should be completed with the information required, i.e. with the content, activity and transport index. The criteria for determination of the category of the package are given in the Table 9.2 below: Table 9.2: Category of package, radiation level and transport index Category

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I WHITE II YELLOW III YELLOW

Maximum radiation level at the external surface of the package, mSv/h, (mrem /h)

Transport Index

0.005 (0.5) 0.5 (50) 2.0 (200)

0 1 10

Transport of Radioactive Materials Both the limits should be satisfied for a package to belong to a specified category. If either of the limits is exceeded, the package would belong to the next higher category. CAUTION: If either the radiation level on the surface of the package is more than 2.0 mSv/h or Transport Index is more than 10, the package should not be forwarded for transportation without prior permission of the competent authority. The Category I WHITE label background color must be white, the color of the trefoil and the printing must be black, and the color of the category bar must be red (Fig. 9.2).

Fig. 9.2: Category I WHITE label (For color version see plate 3)

In the case of Category II YELLOW label, the background color of the upper half of the label must be yellow and the lower half white, the color of the trefoil and the printing must be black, and the color of the category bar must be red (Fig. 9.3). In the case of Category III YELLOW label, the background color of the upper half of the label must be yellow and the lower half white, the color of the trefoil and the printing must be black, and the color of the category bar must be red (Fig. 9.4).

Fig. 9.3: Category II YELLOW label (For color version see plate 3)

Fig. 9.4: Category III YELLOW label (For color version see plate 4)

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Textbook of Radiological Safety PLACARDS Large freight containers carrying packages other than excepted packages, and tanks must bear four placards as specified in Fig. 9.5. These placards must be affixed in a vertical orientation to each side wall and each end wall of the freight container or tank. Any placards which do not relate to the contents must be removed. Instead of using a label and a placard, it is permitted as an alternative to use enlarged labels, with minimum dimension of 25 cm.

Fig. 9.5: Placard, the minimum dimensions is 25 cm. The figure 7 must not be less than 25 mm height. Background colour of the upper half must be yellow and the lower half is white, trefoil and the printing must be black (For color version see plate 4)

If the freight container is packed with radioactive material comprised of a single United Nations commodity, the appropriate United Nations number (Table 9.3) for the consignment must also be displayed, in black digits not less than 65 mm height, either in the lower half of the above placards or in the placard shown in Fig. 9.6.

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Fig. 9.6: Placard for separate display of united nations number. The background color must be orange and border and the united nations number must be black. The **** denote the space for UN number of radioactive material (For color version see plate 4)

Transport of Radioactive Materials Table 9.3: United nations number and proper shipping name UN No.

Proper shipping name and description

2910 2911 2909

Radioactive material excepted package-Limited quantity of material Radioactive material excepted package-Instruments or articles Radioactive material excepted package-Articles manufactured from natural Uranium or depleted Uranium or Natural Thorium. Radioactive material excepted package-Empty packaging Radioactive material, Low specific activity (LSA-I) Radioactive material, Low specific activity (LSA-II). Radioactive material, Low specific activity (LSA-III) Radioactive material, Surface contaminated objects (SCO-I or SCO-II). Radioactive material Type A package, non special form. Radioactive material Type A package, special form. Radioactive material Type B(U) package. Radioactive material Type B(M) package.

2908 2912 3321 3322 2913 2915 3332 2916 2917

BOOKING, STORAGE, TRANSPORT AND DELIVERY OF PACKAGE The package should not be transported as a personal luggage in a bus or in a shared Taxi or in the passenger compartment of a train or in the passenger cabin of an aircraft. It is always booked as an item of cargo. The package should not be dispatched by post. The package is declared as a radioactive consignment in the transport documents. For road transport, the consigner shall declare the consignment by its proper shipping name. The package is provided with transport documents, which include (i) Consigner’s declaration, (ii) TREM card (Transport emergency card), (iii) Instructions to the carrier and (iv) Instructions about emergency measures in case of transport incidents. The key of the lock should be sent along with the transport documents to the CONSIGNEE. The package should not be dispatched with out prior permission of the COMPETENT AUTHORITY. The CONSIGNEE should be informed before dispatching the package and ensured that the CONSIGNEE is prepared to receive the consignment. The CARRIER is provided with documents entitled “INSTRUCTIONS TO THE CARRIER”, while booking the package for transport. The CONSIGNOR, CONSIGNEE and CARRIER should contact the competent authority immediately in the event of : a. Any untoward incident /accident during transport b. Non delivery of the package to the destination within the normal period. It should be ensured that the CONSIGNEE has received the consignment or not and the same is informed to the competent authority. A Check list should be filled to ensure that all the requirements for safe transport of 257 radioactive material are complied with.

Textbook of Radiological Safety CONSIGNOR’S DECLARATION This is to certify that the package containing radioactive material as identified by the following details is safe for transport by rail, road, sea or air. Package forwarded by (consignor) Package addressed to (consignee) Proper shipping name

Radioactive material ————— package, Non fissile

UN class of dangerous goods

7

United Nations No.

UN NO.

Subsidiary risk

Nil

Name of the radioactive material Quantity/Activity of radioactive material

—————RMM on ———-——

Packages details Dimensions of package Weight of the package Type of package* Radiation level on the surface of the package in mSv/h Transport index of the package Category of the package *In the case of Type B(U)/(M) package, competent authority identification number should also be given. I hereby declare that the contents of this consignment are fully and accurately described above by the proper shipping name and are classified, packed, marked and labeled and are in all respects in proper condition for transport according to the AERB Safety code, AERB/SC/TR-1,currently in force.

Date:

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Signature: Name and Address:

Transport of Radioactive Materials TREMCARD Cargo

In-dispersible radioactive material

Nature of Hazard

Radioactive material, Potential external exposure

Emergency action

1. Inspect the package visually. If it is intact, ensure onward journey in the same or another vehicle. 2. In case of fire, fight from a distance 3. If the package appears to be damaged cordon a distance of 3 m around the package. 4. Obtain the names and addresses of persons who might have been exposed to radiation and convey the particulars to the Head, AERB and to the Head, RP and AD, BARC, Mumbai.

Contact telephone numbers for advice and assistance

a. Contact the consignor at the address given on the package b. Chairman, Crises Management Group, DAE, Mumbai-400001, Tel : 022 22023978, 22830441, FAX: 022 228304441 c. Head, Radiological Safety Division, AERB, Niyamak Bhavan, Anushatinagar, Mumbai-400 094, Tel:022 25990655, FAX:022 25990650 d. Head, Radiological Physics and Advisory division, BARC, CT and CRS, Anushaktinagar, Mumbai400094, Tel:022 25519209, FAX: 022 25519209

Telegram

REGATOM,CHEMBUR or HEAD,RP and AD, BARCCHEMBUR

FAX

022 25583230 (AERB),022 255055151(BARC)

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Textbook of Radiological Safety INFORMATION TO CARRIERS 1. The package should be transported by the most direct route. 2. Intermediate off-loading and reloading of the package should be avoided. 3. Package should be handled carefully. Suitable mechanical means should be deployed for handling packages weighing more than 30 kg. 4. Persons should not be allowed to sit on the package or spend more time than the necessary time in the vicinity of the package. 5. The package should not be transported along with other dangerous good such as explosives and inflammables. 6. The package should not be transported/stored together with photosensitive films/plates. 7. The package should be kept segregated from spaces occupied by passengers and public. 8. If several packages containing radioactive material are to be transported, then the total number of packages loaded in a single vehicle should not be so restricted that the sum of the transport indexes of the packages does not exceed 50, except in case of executive use. Further the total number of packages staked in a storage area should be so limited that in a given stack the above limit of 50 of the sum of transport indexes is not exceeded and such stacks containing radioactive consignments are separated by at least 6 meters. 9. If the shipment is under explosive use, i.e. the entire conveyance is for the proposed transport of radioactive material then (a) there should not be any intermediate loading and unloading operations of other goods. (b) Nothing other than the intended radioactive material along with its accessories should be carried in this vehicle. 10. At the destination, it should be ensured that the package is delivered to the consignee to whom it is indeed addressed. 11. One copy of the TREMCARD should be carried in the vehicle carrying the radioactive cargo. If the package(s) get (s) involved in an accident or get (s) damaged during transport, the instructions specified in the TREMCARD should be implemented. 12. If the package is not claimed by the consignee at the destination, it should not be auctioned or otherwise disposed of. The matter should be brought to the notice of the consigner and Head, RSD, AERB, Niyamakbhavan, Anushaktinagar, Mumbai-400094 and such measures as recommended in this regard by HEAD, AERB, Mumbai, should be duly implemented.

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Transport of Radioactive Materials Annexure-I Check list for preparing, marking and labelling a package to transport radioactive material Type of the package:

Type A /Type B (U)/Type B (M)/ other (please specify)

If Type B(U) or Type B(M), give the Competent Authority Identification No. _________ Preparation of the package 1. Whether it was confirmed with radiation survey meter that the radiation source is in its proper storage place in the shielded container/source housing/ radiography camera. Yes/No 2. Whether the source is locked/arrested in its shielded position. Yes/No 3. Whether all the nuts and bolts meant for fastening the shielding are properly in place, secured and tightened. Yes/No 4. Whether the shielded container is properly immobilized in the outer container/ box (package) with the help of wooden spacers, etc. Yes/No 5. Whether the outer box (package) is properly closed with the help of fasteners/ bolts, steel straps etc. and the nuts/bolts are properly tightened. Yes/No 6. Whether the package is properly locked and sealed with crossed steel strips. Yes/No Markings of the package 1. Whether the Gross weight of the package is marked on it, if the weight exceeds 30 kg. Yes/No 2. Whether the package is marked on the out side with name, address and telephone number of consignor (sender) and consignee (receiver). Yes/No 3. Whether the shielded container inside the package is marked with name, address and telephone number of consignor (sender) and consignee (receiver). Yes/No 4. Whether the proper shipping name of the radioactive material is marked on the package. Yes/No

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Textbook of Radiological Safety 5. Whether the proper United Nation Number is marked on the package. Yes/No 6. Whether the Type of package is marked on it. Yes/No 7. Whether all the markings made on the package are legible and durable. Yes/No Labeling 1. Whether label of proper category is selected based on the radiation. Yes/No 2. Whether the package is labeled with two numbers of selected category labels affixed on two opposite sides of the package. Yes/No 3. Whether these labels are properly filled-in with respect to (a) Name of Radio nuclide (b) Activity in becquerel (c) Transport Index. Yes/No Transport documents 1. Whether the Consignor’s declaration along with particulars of the consignment are provided in proper format. Yes/No 2. Whether a copy of “instructions to the carrier” is provided to the carrier and the carrier is properly informed regarding the radioactive nature of the consignment and the hazards associated with it. Yes/No 3. Whether the TREMCARD is provided to the driver of the vehicle and whether it forms a part of the Transport Document. Yes/No 4. Whether the copy of “Emergency instructions in writing” is provided to the driver of the vehicle. Yes/No Prior to actual transport of the package 1. Whether the carrier is informed that the package should not be carried in the passenger compartment of a train, an aircraft or passenger cabin of a ship or in a passenger bus or a shared taxi or any shared rented vehicle. Yes/No 2. Whether the carrier is informed about the care to be taken during handling and carriage of the package in trans-shipments, etc. Yes/No

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Transport of Radioactive Materials 3. Whether carrier is informed about the proper stowage of the package in the vehicle during the transport. Yes/No 4. Whether the carrier is informed that the package should be immobilized during the transport. Yes/No 5. Whether the consent of the consignee is obtained before dispatching the package. Yes/No

Consignor’s Signature: Name and address Date: Seal.

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Textbook of Radiological Safety Annexure-II Instructions in writing regarding Practical Measures for Transport Incidents Involving Radioactive Cargo (AERB/RSD/TRANSPORT EMERGENCY/REV.6)

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1. About the package: Packaging which are permitted to be used for transport of radioactive materials are generally designed to prescribed standards aimed at prevention of release of the contents and of excessive exposure of public of radiation. Essentially there are two types of packages, namely, Type A and Type (B)U/(M). If a Type A package is involved in an accident which may result in the package falling off the vehicle, it is very unlikely that the package will be broken open. If the accident is severe such as vehicle rolling over, then the package may be damaged through the loss of shielding or release of the contents may not occur. A Type B package is designed to withstand severe accidents Only those packagings whose design and specifications have been duly approved by Atomic Energy Regulatory Board (AERB) for Type B(U)/(M) and registered in AERB for industrial and Type A package, are deployed for the transport of radioactive material in India. It is such a package which is loaded in the vehicle. 2. Nature of Hazard: The hazard associated with radioactive consignment is exposure to radiation. Such exposure may be external and/or internal in nature. If the radioactive content is an in dispersible solid or capsule, the hazard is likely to be external. If the content is in dispersible form, in the unlikely event of a severe accident, the potential for internal and some times, in addition, external exposure may exist. 3. Protective devices to be carried in the vehicle: The driver of the vehicle and his assistant should each have some protective device if the vehicle carries a package containing dispersible radioactive material. • The protective equipments include, • Two pairs of rubber shoes • Two pairs of latex gloves • Coveralls, 2 numbers • Big empty polythene bags: 6 numbers • Big (3 m x 3m) polythene sheets • One kg of cotton wool. If the vehicle does not carry any package containing dispersible radioactive material the protective equipment would not be required from radiation safely standpoint. 4. Emergency action and first aid: If an accident occurs, don’t panic. Rescue the injured. If life is at stake, save life. It is unlikely that in a transport accident involving the commonly deployed small Type A and Type B(U)/(M) packages any significant injury to the rescuer will result from radiation. If any of the packages which are damaged in the accident was containing radioactive material

Transport of Radioactive Materials in a dispersible form, which could not be breathed in, hold a cloth towel or a handkerchief over your mouth and nose. If there is fire, summon assistance from the local public and fire brigade. Fight fire from a distance. Follow these instructions: • Fight fire as far upwind as possible • Keep out of smoke, fumes and dust • Wear the coverall, gloves and shoes and cover mouth and nose with handkerchief • Spend minimum time near the package • Keep by standers upwind at least 5 m away Inspect the packages. If the packages appear to be intact, ensure onward journey in the same vehicle. If the vehicle cannot be release for onward journey for a long time, then arrange for onward journey of the package in some other vehicle. If the package appears to be damaged, wrap it in a polythene bag, segregate the package and cordon a distance of 5 m around the package. If the contents of the package appears to have spilled, then take the following measures: • Assume that the area and the objects on which the spillage has occurred are contaminated. • Wear the shoes, gloves and coveralls • Collect the spillage, using cotton wool, in a polythene bag • Wrap the damaged package in polythene bags • Cover the contaminated objects and contaminated area with polythene sheets. • Do not eat, drink or smoke within the cordon • Take measures to prevent a fire accident. • Seek assistance from AERB/BARC as directed in para 5 below • Do not allow the public within the cordon unless so advised by the radiological safety authorities from AERB/BARC, Mumbai. All persons who were engaged in the emergency response measures should carefully and thoroughly wash the affected parts of the skin with plenty of water. Obtain the names and addresses of persons who may have been exposed to radiation and convey the particulars to the Head, RSD, AERB, Niyamak Bhavan, Anushaktinagar, Mumbai – 400 094. 5. Telephones for advice and assistance for advice and assistance contact: Chairman Crisis Management Group, DAE Mumbai–400 001 Tel.(round the clock) 022-22023978, 22830441, Fax:022-22830441. Head, Radiological Safety Division, AERB Niyamak Bhavan, Anushaktinagar, Mumbai-400 094, Tel.(off) 022-25990655, 27824986 (Res), Fax:022-25990650

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Textbook of Radiological Safety Head, Radiological Physics and Advisory Division BARC, CT and CRS, Anushaktinagar Mumbai-400 094, Tel. 022-25519209 (Off), 022-25517812 (Res), Fax: 022-25519209 While seeking advice and assistance furnish the furnish the following particulars: • The place where the accident occurred. • The date and time of occurrence of the incident. • Whether the incident involved impact, fire or both. • Details of emergency action taken. • The condition of the packages, whether damage/Spillage suspected. • The name and addresses of persons who may have been exposed to radiation. Act exactly in accordance with the instructions given by the above authorities. Onward journey of the packages which were damaged in the incident may be arranged only after obtaining clearance from the above authorities. 6. General: Every driver should ensure that he is completely familiar with the “Instructions in Writing…..” and the procedures recommended in the TREMCARD. Prior to undertaking the journey, the driver should ensure that he carries the following items with him: the ”Instructions in writing…..” the “TREMCARD” the protective devices as specified in para 3 above. The assistant accompanying the driver should also be familiar with these instructions.

BIBLIOGRAPHY 1. AERB safety code No.SC/TR-3: Emergency response planning and preparedness for Transport accidents involving radioactive material. 2. AERB Safety code No.SG/TR-3:Procedure for forwarding, transport, handling and storage of radioactive consignments. 3. AERB safety code: No.SC/TR-1:Transport of radioactive materials

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Chapter

10

Radioactive Waste Disposal

INTRODUCTION Every industry generates some amount of waste and nuclear industry is one among them. The waste generated in the nuclear industry is called radioactive waste, which may be in soild, liquid or gaseous form. Hazards related to radioactive waste give rise to certain amount of fear and unacceptability in the minds of public. The radioactive waste needs to be managed safely to ensure protection of man and environment, without imposing significant burden on future generations. If not handled carefully, ionizing radiations emitted by the radioactive waste can cause somatic and genetic effects in the living beings. Radioactive waste can be treated some extent by physicochemical methods, adopted to conventional pollutants. However, they undergo decay and the radiation comes to the background level after a certain period of time. This period depends upon the half life of the waste, which vary from seconds to thousands of years. Hence, effective waste management methods are the need of the hour. WASTE MANAGEMENT The basic objective of radioactive waste management is: i. Protection of human health, ii. Protection of environment, and iii. Protection of future generation. To achieve this the methods adopted in the practice includes: i. Minimize the generation of radioactive waste, ii. Recycling and reuse the waste material, and iii. Minimize the exposure to operation staff and public. The basic approaches used in the management of radioactive wastes are: i. Delay and decay ii. Dilute and disperse iii. Concentrate and contain. The delay and decay is suitable only for short half life isotopes. For example I-131 (HL-8 days) in small volumes may be retained till the activity levels comes down to the desired values, suitable for release into the environment. Where as the later two methods are generally adopted in the management of all radioactive wastes.

Textbook of Radiological Safety Delay and Decay It is based on the fact that radionuclides lose their radioactivity through decay, and this fact may be utilized in the treatment not only of intermediate and high level solid, liquid and gaseous wastes but in some cases also in that of low-level wastes. The aim is to ease problems in subsequent handling or to lessen risks of releases to the environment, taking advantage of the decay of some radionuclides – particularly those having short half lives – with the passage of time. The principle is especially useful for those installations where a substantial reduction in the activity level of a waste stream can be achieved by delaying discharge of effluents for a few days. Dilute and Disperse The principle of dilution and dispersion is based on the assumption that the environment has a finite capacity for dilution of radionuclides to an innocuous level. The application of this principle requires an understanding of the behaviour of radioactive materials in the environment and of the ways in which the released radionuclides, particularly those that are considered to be critical, may lead later to the exposure of man. It is especially important to take into consideration environmental processes which may cause reconcentration of radionuclides. Concentrate and Contain The principle of concentration and containment derives from the concept that the majority of the radioactivity generated in nuclear programs must be kept in isolation from the human environment. Since some radionuclides take a long time to decay to innocuous level, some wastes must be contained for extended period of time. The principle is invoked in techniques for air and gas cleaning; the treatment of liquid wastes by scavenging and precipitation; ion exchange and evaporation; the treatment of low-level, solid wastes by incineration, baling and packaging the treatment of intermediate-level solid and liquid wastes by insolubilization in asphalt; conversion of high-level liquid wastes to insoluble solids by high-temperature calcinations or incorporation in glass; tank storage of intermediate – and high-level liquid wastes; storage of solid wastes in vaults or caverns; and disposal of solid and liquid wastes in deep geological formations. SOURCES AND NATURE OF WASTE Mainly radioactive waste is generated from: i. Nuclear fuel ii. Research and power generation reactions, and 268 iii. Isotope applications.

Radioactive Waste Disposal In nuclear fuel it begins with the mining, milling, refining of U and Th from their ores, fuel fabrication and fuel reprocessing. The second type includes isotope production and electricity generation. The third, isotope application includes medicine, nuclear research, industry and agriculture. The waste generated may be liquid, solid or gaseous with radioactive content varying from insignificant to extremely high levels. Wide range of radionuclides are used in medicine for treatment, diagnosis and research. The waste arises from the above applications are called medical radioactive waste, which require a comprehensive management system. The type of radioactive waste that may arise in medicine are mainly from radiotherapy and nuclear medicine. The sources from which the wastes arises are source storage, preparation, source administration and injection needles, counting room, toilet-improper use (vomiting by patients) and decontamination. In radiothepay the waste arises from; i. Used Tele-cobalt -60 sources, enriched uranium components (shutter, collimator etc.), and ii. Used Brachytherapy sources (Co-60, Cs-137, Ir-192, and I-125), leaking sources and contaminated cotton, etc. In nuclear medicine the radioactive waste arises from: iii. Decayed sealed sources iv. Spent radionuclide generators (99mTc,81mKr,and 185mAu etc.) v. Laboratory solutions of low activity vi. Low activity liquid washings from vials vii. Liquid scintillants immiscible with water viii. Biologically contaminated solid waste (e.g. syringes, vials) xi. Radioactive gases. CLASSIFICATION OF WASTE The radioactive waste is classified as solid, liquid and gaseous. The liquid and gaseous wastes are further categorized on the amount of radioactivity. The solid waste is categoried, depending on the radiation dose on the waste package. In India, conforming with the internationally acceptable norms and standards, the Atomic Energy regulatory Board (AERB) has categorized these wastes, as given the Table 10.1. Table 10.1: Classification of radioactive waste Category Solid Surface dose, mGy/h

Liquid Activity, Bq /ml

I II III IV V

<3.7 × 10-2 3.7 × 10-6 -2 1 3.7 × 10 to 3.7 × 10 3.7 × 10-6 to 3.7 × 10-2 3.7 × 101 to 3.7 × 103 >3.7 × 10-2 3.7 × 103 to 3.7 × 108 >3.7 × 108

<2 2-20 >20 Alpha bearing -

Gaseous Activity, Bq /ml

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Textbook of Radiological Safety The liquid waste of categories III and IV need to be properly shielded whereas category V requires shielding as well as cooling during handling, storage, treatment and disposal. The classification signifies the basic requirements for safe handling and disposal of wastes. The term low level waste (LLW), intermediate level waste (ILW) and high level waste (HLW) are also employed in normal practice and for day to day working. Waste trace levels to fractions of mCi comes under low level waste. Where as, 1mCi-1Ci / litre and 1Ci-100 Ci / litre comes under intermediate and high level waste category. These terms give an idea about magnitude of the radioactive contents and the associated radiation. The ICRP-25 classification of nuclear medicine laboratory is given in Table 10.2 Table 10.2: Classification of nuclear medicine laboratories in terms of activity (ICRP-25) Classification Group 2: 125 131 I, I Group 3: 201 Tl, 32P, 51Cr, Group 4: 99m Tc, 133Xe

99

Mo

*supervised areas

Low

Medium

High

< 500 kBq,

500 kBq-500 MBq

500 MBq-5 GBq

< 5 MBq

5 MBq-5 GBq

**5 GBq-500 GBq

< 500 MBq

*500 MBq-500GBq

500 GBq-50 TBq

** Generator room

TYPES OF RADIOACTIVE WASTE Liquid Waste Liquid waste includes contaminated water and effluent, waste arising from chemical processing and decontamination solutions, solvents, blood or body fluids, discharged liquid radiopharmacheuticals, wound and oral discharges, urine etc. The waste that includes both radioactivity and a hazardous chemical component is referred as mixed waste. A safe disposal is one in which no member of the public should get more than the effective dose limits. The liquid discharge systems should be 10-4 to 10-5 Ci / ml. Low and intermediate active liquid waste from different sources is normally collected and transported to the treatment facility by means of permanent pipelines systems. In some cases, where volume involved is small, specially designed tankers are used for collection and transportation. A wide variety of treatment methods are available to meet specific requirement of decontamination. Decontamination factor is defined as a ratio of radioactivity content of untreated and treated waste. The commonly employed processes and the corresponding decontamination factors are 270 given in Table 10.3.

Radioactive Waste Disposal Table 10.3: Decontamination factors for various processes Processes

Decontamination factor

Chemical precipitation Ion exchange Reverse osmosis Evaporation

10-100 10-10,000 10-50 1000-10,000

Chemical Treatment In this insoluble flocs of phosphates, sulphates, hydroxides and complex metal ferrocyanides are used to remove radionuclides from the waste. Certain selected chemicals such as calcium or barium chloride, trisodium phosphate or sodium sulphate, potassium ferrocyanide, copper sulphate and ferric ion are mixed with the effluents in predetermined quantities at an optimum pH value. The resulting precipitate flocs incorporating radioactivity are allowed to settle and are separated from the supernate liquids depleted in radioactive content. The sludges are further concentrated and dewatered by filtration or centrifuging. The resulting solids have highly concentrated activity and are subjected to further processing before disposal. Ion Exchange This is the technique used for removal of specific radionuclides from the bulk of wastes. Naturally occurring ion exchange materials like vermiculite and bentonite are most commonly used for this purpose. Synthetic ion exchangers are also used for the decontamination of the waste. These are specially useful for both clean liquids as well as those containing high percentage of dissolved salts. Mobile transportable ion-exchange system is also in use. Reverse Osmosis It is an important process used in the decontamination of low and intermediate level liquid wastes. It employs membranes like polyamide and pressure of the order of 20 kgcm2. The waste is pretreated for pH adjustment and then filtered for removal of complex agents. The membrane separates the waste into two components, reject and permeate. The volume of waste is normally reduced by a factor of 10 by this process. If required, the concentrate is further treated by evaporation, prior to its solidification. Evaporation It is used for concentrating the liquid waste. Steam and natural evaporation methods are employed, depending upon the activity, volumes involved and climatic conditions. For intermediate level and high level waste, steam evaporation is preferred whereas for large volumes involving low activity, 271

Textbook of Radiological Safety natural evaporation is desirable. Very high volume reduction with practically zero release of activity is possible by use of non boiling solar pans. Solar evaporation process is efficient for tritiated water. Solid Waste Solid waste is generated at different stages in many different forms which include tissue papers, plastics, contaminated materials, discarded containers, protective wears, worn out metallic parts and equipment and accessories, spend radiation sources etc. Solid radioactive waste also consists of general biomedical waste, that includes protective clothing, plastic sheets and bags, gloves, masks, filters, overshoes, paper wipes, towels, metal and glass, hand tools and discarded equipment. Disposal of Radioactive Solid Waste The waste material is segregated into: i. Compressible and combustible, and ii. Non-compressible and non-combustible waste. The solid waste material can be concentrated and disposed by the following methods: i. Incineration ii. Burial of solid waste in the ground iii. Storage of protected material iv. Sea dumping. Up to 2 mCi can be disposed as ordinary waste. Occasional release up to 100 mCi in ordinary dust bin is allowed. Sealed sources > 10 mCi should not be disposed and they must be kept for decay. Higher activities can be buried or burnt and short lived activity up to 1mCi can be burnt. Incineration Incineration will substantially reduce the volume of wastes, but the total radioactive content will not be reduced. Depending upon the physical and chemical characteristics of the compound involved, the activity may be deposited in the gaseous effluents, inside surface of the incinerator and in the ash. The activity associated with the incinerated waste must be restricted to the public exposure limits. This method is suitable in reducing the volume of the waste with little escape of activity to the environment. The incinerators are specially designed to remove the radioactivity from combustion gases by the use of scrubbers and filters and to ensure that the radioactive ash is contained so as to not to cause an airborne hazard.

272

Radioactive Waste Disposal Burial of Solid Waste Any site chosen for such burial operations should be examined carefully with regard to its geologic and hydrologic properties and an assessment should be made of the possible contamination of water supplies and of ecological systems that might lead to human exposures. The depth chosen for burial must be sufficient to prevent leakage of any harmful level of radioactivity into usable surface waters or ground waters. The whole of the burial area and its immediate vicinity should be isolated suitably and fenced to prevent use of the area. A central record should be maintained of all burials, inorder to assure that the area is kept under continuing surveillance. Under such conditions some loaching of the radioactivity into the ground water can take place, and this may therefore be considered under the principle of dilute and disperse to a small extent. a. Unprotected material: The following method is suitable for low level radioactive waste. The burial must be carefully controlled to avoid spread of contamination to surface ground waters. The choice of burial site is very important and it should be suitably cordoned off. The solid type is also important since fixation on a good exchange material (e.g.montmorrillonite), will reduce the movement of activity in the ground. In this method one should ensure that the radioactivity is in the form of completely insoluble material. The size of the pits is 120 cm × 120 cm and depth of the pit should be such that there should be 120 cm of earth above. The amount in the pit should be restricted and the minimum distance between the pits should be 180 cm. Not more than 12 burial pits per year and after 7-8 half lives, the pits can be empted into the municipal dump and reused. A record should be maintained about the location of each pit, the identity and quantity of each radionuclide buried in it. The maximum disposal limits for ground burial is given in the Table 10.4. b. Protected material: Highly radioactive materials can also be buried in the soil if a material is treated in such a way that no loaching of the material can occur. In general, this involves packing the waste material in steel drums or concrete blocks. In the case of very high radioactive materials, a concrete tomb is constructed underground and the material fed into the tomb via a chute from the surface. After the tomb is filled up the whole construction is filled with asphalt and sealed. Municipal dumps can be used for protected burial, because it is normally ensured that they carefully sited to prevent contamination of surface or other waters with conventional non-radioactive pollutants. The radioactive materials should be packed in steel bins and then must be buried beneath at least 1.5 meters of other rubbish. Up to 1 mCi can be buried in a municipal site in a form unattractive for salvage.

273

Textbook of Radiological Safety Table 10.4: Disposal limits for ground level Radionuclide

Maximum activity in a pit (MBq)

H C 24 Na 32 P 35 S 45 Ca 59 Fe 99 Mo 125 I 131 I

9250 1850 370 370 1850 370 370 370 37 37

3

14

Storage The purpose of protective storage is: i. To avoid dust hazard, ii. To provide shielding, and iii. To prevent accumulation in working areas. Normally short lived isotopes are stored for decay. Long lived radioisotopes are stored and disposed to centralized waste management facility, BARC, Kalpakkam or Mumbai. There the waste is disposed off in engineered structures such as reinforced concrete trenches and the tile holes depending upon the waste and the radioactivity. In the case of Brachytherapy solid wastes, it is stored and then returned to the supplier, which is the procedure adopted in the country as on date. The storage area should be in accessible to unauthorized persons. Usually foot operated dust pins with plastic bag is used in hospitals to store solid waste. Proper labeling of the container along with records are mandatory. The radiation level at 1 meter should not exceed 1.5 mR/h in such storages. Sea Dumping

274

This method of disposal for solid active waste can be carried out after carefully choosing an area on the basis of oceanographic studies. For low activity solid waste use is made of relatively shallow waters of approximately 100 fathoms. This method is practiced in the U.K and in the US both packed and unpackaged wastes are dumped in to water exceeding 1000 fathoms in depth. Containers used for sea dumping should be designed, constructed and filled in such a manner as to ensure the following: i. That they can not be easily damaged or broken and will reach the bottom with out appreciable loss of contents; ii. They are free from voids; iii. They have density of 1.2 g/cc;

Radioactive Waste Disposal iv. They are provided with sufficient shielding for safe storage; v. They are of a size and shape to be handled quickly and conveniently. Although no general legislative control exists for the dumping of material outside territorial waters, it is highly desirable that careful records are maintained of the total activity content and weight of all consignments. Gaseous Waste Gaseous waste management is very important to take care of airborne radioactive particles and gases. The main contaminants of importance in nuclear facility are radio-iodine, tritium, fission products, noble gases etc. Once the effluent is released in the air the operator has no control and hence can not escape the consequences arising out of air pollution. The emission of activity to the atmosphere may give rise to three possible types of hazard: (i) a direct irradiation hazard from the radioactive clout itself or from material which is deposited on the ground, (ii) inhalation hazard to people breathing the cloud, and (iii) an ingestion hazard from material that finds its way into food chains. The type of hazard depends on the circumstances of a particular emission. For most isotopes the hazard is usually caused either by first or third types. Therefore the airborne activity in the working area should be kept within limits (Table 10.5). The removal of particulate and gaseous contaminants from gaseous effluents is a complex and expensive one. It is advisable to design the plant and buildings so that the volume to be treated is as small as possible. This may be achieved by providing separate ducking systems for radioactive and non radioactive effluents and filtration system for the radioactive fraction alone. In addition, negative pressure compared to atmosphere is maintained in the working areas to restrict the release of activity to the environment. The ventilation exhaust system is provided with suitable devices to contain airborne radionuclides. The exhaust gases are treated for removal of and retention of particulate activity by using high efficiency particulate air (HEPA) filters along with other cleaning techniques. Table 10.5: Air concentration that would result annual dose limits to occupational workers Radionuclide

Air concentration borne (μCi / ml)

F Tc 131 I 14 C 133 Xe

3 × 10-5 6 × 10-5 2 × 10-8 1 × 10-6 1 × 10-4

18

99m

275

Textbook of Radiological Safety DISPOSAL OF LOW ACTIVITY WASTES INTO THE ENVIRONMENT The principle involved is dilute and disperse. Disposal of diluted waste can be done into the sea, river or into the ground. It will depend upon the proximity of the facilities producing or treating the wastes to the types of environment mentioned aforesaid. Sea is a unique medium for disposal of low activity waste because dilution, factors are large and maximum permissible levels for discharge are higher than those for fresh water bodies, as drinking water tolerances are not involved. However, dilution may be partly off set by concentration of some radioactive species in marine life which may enter into feed cycles ending in consumption by human beings. Hence, any projected sea or river disposal scheme must be preceded by extensive trails to determine the dilution of foreign solution discharged at various points in the sea or river. By these the degree of horizontal and vertical mixing is obtained on discharge to the sea or river as well as the effects of winds and tides upon the dispersal of the material. Detailed biological experiments must also be done to determine the pattern of up take of radioactivity by marine fauna and flora, especially those involved in food cycles. For the disposal of liquid wastes into rivers the following factors should be considered: flow rate, nature of river bed, turbidity, currents etc. Consideration must be given to the types of water utilization such as drinking, industrial and agricultural uses and appropriate correction factors should be applied to arrive at the permissible discharges. Dilution of lowlevel liquid wastes can be achieved by a. Addition of uncontaminated liquid to reduce the concentrations prior to discharge. b. Release of the liquid wastes at small rates over long periods of time. c. Release of wastes into large bodies of water. DISPOSAL OF RADIOACTIVE EFFLUENT INTO THE GROUND The liquid waste injected into the ground will tend to percolate downwards until it reaches the ground –water table, when it begins to travel forward in the direction of underground water flow; in doing so, the activity is diluted by the natural water flowing in the ground. In addition, if the permeability and porosity of the formation are such that underground water is appreciable in volume but not too rapid in rate, the combination of dilution and the decay afforded by an appreciable retention time may be such that the waste levels are below the dose equivalent limits, by the time the waste reaches the environment, by discharging to springs, streams or the sea. The retention time may be extended by chemical reactions between the waste and the soil, and in very favorable circumstances very high retention 276 times may be encouraged. The factors determining the applicability of

Radioactive Waste Disposal ground disposal methods are two, namely (i) hydrological, and (ii) chemical. The desirable factors for hydrological factors for ground disposal can be summarized as follows: i. A deep water table with good flow gradients. ii. Appreciable permeability to allow rates of flow sufficient to be useful. iii. A fairly high porous thick formation to restrict rates of flow. iv. Wide spacing of water bodies such as lakes and streams so that long distances of underground flow are involved. v. Relatively low rainfall. Although the ground water flow may then be slight, the porosity of unsaturated formation in an area of low rainfall may afford high retention volumes. In addition to the hydrological phase retention in the ground there may be further retention of certain species due to chemical reactions with constituents of the soil or rock. These reactions are precipitation and ion exchange. The precipitation characteristics will depend upon the natural pH of the soil. Ion exchange properties of soil depend upon the type and amount of clay minerals present, and in soils rich in organic matter to the humus content as well. In any case ground disposal requires detailed study of hydrological and chemical characteristics of the soil where the disposal is contemplated. DISPOSAL OF P-32 AND I-131 INTO MUNICIPAL SEWERS BY MEDICAL USERS The various waste disposal methods are (i) toilet disposal, where the patient administered with radioactive isotopes is allowed to use the toilet without restriction, (ii) Batch bottle disposal, where the radioactive material is collected in a 5 litre bottle, diluted to the top and poured into the sink, and (iii) constant drip discharge: In order to maintain a uniform discharge in case where the activity to the disposed of is more than 10 mCi of P-32 and I-131 a day for 4.5 million litres of sewage flow, it is necessary to use a constant drip discharge bottle. As a rule of thumb, up to 100 mCi of P-32 or I-131 may be discharged through a constant drip discharge bottle during a 6 hour day light period, when the flow in a sewage system is 4.5 million litres during dry weather. The limit is subject to revision depending upon variation in flow rate and actual radioactive measurements in the sledges. Any waste disposal method must take into account the following: i. Permissible concentrations applicable from the standpoint of community safety, especially with regard to sanitation workers and sewage plant personnel (Tables 10.6 and 10.7). ii. To formulate practicable rules and activity discharge based on the average water consumption and average isotope concentration level arising there from. iii. To ensure that the degree of dilution envisaged will be at the discharge point from the institution into the sewage system into which the 277 radioactive wastes are discharged.

Textbook of Radiological Safety iv. To bear in mind that it would be unreasonable to insists upon the dilution of radioactive waste in the sewage to the level established as maximum permissible limits for drinking water. v. to ensure that the external radiation hazard to sanitation and sewage plant personnel will not be more than that caused by accidental immersion in a concentration of 0.1 mCi / litre of sewage. vi. The hazard to the general population in the event of the sludge containing radioactive material being used as fertilizer. Table 10.6: Airborne, water and sewer concentrations that would result annual dose limits for public Radionuclide

Environmental concentrations (µCi/ml) Air Water

Sewer concentration (μCi / ml)

F Tc 131 I 14 C 133 Xe

3 × 10-9 2 × 10-7 2 × 10-10 3 × 10-9 5 × 10-7

3 × 10-4 1 × 10-2 1 × 10-5 3 × 10-4 —

18

99m

3 × 10-5 1 × 10-3 1 × 10-6 3 × 10-5 —

DISPOSAL OF RADIOACTIVE WASTE FROM NUCLEAR MEDICINE PROCEDURES Radioactive waste from nuclear medicine procedures can be dealt with either by simply storing the waste safely until radioactive decay has reduced the activity to a safe level or possibly by disposal of low activity waste into the sewage system, if permitted by the local regulatory authority. Long half-life or high activity waste may need long term storage in a suitable storage area. Waste materials from the drawing up of patient injections can be divided into two groups, those with long and those with short halflives. Table 10.7: Disposal limits for sanitary sewage systems Radionuclide

Maximum limit on total discharge per day (MBq)

Average monthly concentration of radioactivity in the discharge (MBq / m3)

H C 24 Na 32 P 35 S 45 Ca 99 Mo+99mTc 125 I 131 I

92.5 18.5 3.7 3.7 18.5 3.7 3.7 3.7 3.7

3700 740 222 28.5 74 10.1 185 22.2 22.2

3

14

278

Radioactive Waste Disposal Technetium-99m waste normally requires storage for only 48 hours, in a plastic bag inside a shielded container. The container should be labeled with the radionuclide and date. Gallium-67, I-131 and other longer half-life materials should be placed in a separate labeled and dated plastic bag and stored safely. Sharp items, such as needles, should be separated and placed in a shielded plastic container for safety. When disposing of waste, attention should be paid to the following points: 1. Normally once the surface dose rate in any individual bag of waste is below 5 mGy/h it can be disposed of. 2. Disposable gloves should be worn and caution exercised when handling sharp items. 3. Any labels and radiation symbols should be removed. 4. Waste should be placed in a locally appropriate waste disposal container, for example, a biological waste bag. 5. Formulate practicable rules and discharge levels based on average water consumption and isotope concentration. 6. The radiation hazard to sanitation workers and sewage personnel < that from accidental immersion of 0.1 mCi / litter. 7. The effective dose limits should not be exceeded. 8. If the effluents is used for agricultural purposes the levels should be reduced by 10. Possible build up of activity in irrigated land and crops to be considered. 9. Total activity released should not exceeded 1 Ci / year. 10. Long lived isotopes other than 3H and 14 C should not be released to sink. Each type of waste in nuclear medicine requires special consideration since biological contamination (e.g. blood) may be a more series hazard. Accepted levels for disposal of radioactive waste under controlled and non controlled conditions are given in Table 10.8. A controlled disposal is defined as disposal with permission from the regulatory authority. Records should be kept listing initial activities and recommended disposal dates for medium and long half life nuclides.99mTc waste should be kept for an appropriate decay period before disposal (24h); no record normally required for decayed 99mTc contaminated items. Table 10.8: Discharged activity limits for non controlled and controlled conditions Classification

Non controlled (Bq)

Controlled (Bq)

Group 1 Group 2: 125I, 131I Group 3: 201 Tl, 32P, 67Ga, 51Cr, 111 In, 57Co, 58Co, 99Mo 99m Group 4: Tc, 133Xe

— 5 × 104 (1.4 μCi) 5 × 105 (14 μCi)

— 1 × 107 (270 μCi) 5 × 106 (140 μCi)

5 × 106 (140 μCi)

5 × 107 (1.4 μCi)

279

Textbook of Radiological Safety Tc Generators

99m

A Mo-99 generator with activity of 345 mCi decays to 140 μCi after 31 days. It can be disposed under controlled conditions with proper authorization. Spent generators should be removed to a separate store room or bunker for the required decay period. Storage room should be provided with shielding, so that the exposure should not exceed the effective dose limits. Patient Waste Special toilet should be available to nuclear medicine patients. The toilet should have direct access to the sewage system and should not run under the nuclear medicine department, since high activities will affect the performance of counters and imaging devices by increasing background activity levels. Patient’s excreta are exempted from disposal restrictions. Urine and feces should be discharged using a toilet connected directly to a main sewer. ROUTINE PROTECTIVE CLOTHING Laboratory Coat Conventional white cotton drill or nylon coat of proper size which should extend below the knees are suitable for clean areas. Overalls (Boiler suit) These are one-piece cotton drill garments so designed as to cover the body completely except for the head and neck, wrists and hands, and feet and angles. The fastenings for these garments are usually at the front. They are extremely useful as they protect all the clothing worn below. Aprons In areas in which the processes involves work at benches with liquids, an apron of suitable impervious material such as PVC, Polyethylene or neoprene will be found useful in preventing the clothing below from becoming contaminated by corrosive liquids or dust. Rubber Gloves For general laboratory work, surgical gloves are adequate for most operations. Where it is necessary to handle beta active material directly with the hands, rubber gloves of a heavier type or leather gloves may be used to reduce the beta radiation dose to the hands. Foot Wear

280

These should be preferably rubber-soled to prevent the uptake of contamination and to facilitate cleaning. It is recommended that the pattern

Radioactive Waste Disposal of the rubber sole should not be too deeply indented. The upper part of the shoes should be well waxed to resist the absorption of contaminated solutions. Overshoes These are worn over the normal walking shoe and are suitable for use by visitors to active areas or for general use in laboratories. The conventional rubber overshoes are suitable but the soles should not be too deeply indented. A cheaper form of overshoe made of rubber, plastic or canvas is also available. Rubber Boots These are particularly useful for wear in areas in which the processes involve contaminated solutions or wet conditions, such as in areas being decontaminated. The half length rubber boot is usually adequate for this purpose. The soles of these boots should not be too deeply indented. Breathing Apparatus For work in areas of low or medium level of airborne activity, a full face respirator with an efficient filter provides adequate protection. The filter used must be reliable; suggested types are the resin wool and charcoal or the highly efficient paper filters which are commercially available. Care must be taken to ensure that these respirators fit properly and do not allow air to be taken in from the sides of the face-piece. For work in areas of very low activity a half-face respirator may be used. DECONTAMINATION PROCEDURES Decontamination is the process of removal of radioactive contamination from the skin or from surfaces such as the wall or floor of working areas. Radioactive contamination may exist in loose form or may be more or less fixed as a result of physical and chemical factors. Whenever possible contamination should be cleaned up as soon as it occurs. This further prevents the spread which makes the eventual decontamination more necessary. Skin and Surface Contamination In decontaminating the skin, while it would be ideal to remove the entire contamination, this may not always be possible, because the drastic measures which may be necessary in certain causes could result in such damage to the skin that the radioactive material could gain entry into the body and so give rise to an internal hazard. In such cases, it should be considered satisfactory to reduce the levels of contamination to 281

Textbook of Radiological Safety within permissible limits. Similar considerations would apply to the decontamination of surface such as walls, floors, and table tops, and for contaminated equipment. There could, however, be situations in which experimental requirements render it essential that the decontamination be absolute. On the other hand, in dealing with contamination of certain articles and types of equipment, it might turn out to be more economical to store the contaminated object temporarily, with a view of letting the activity die down naturally to within permissible level, or to dispose of it as waste. The above considerations imply the setting up of maximum permissible levels of contamination for the skin and for surfaces in controlled and uncontrolled areas.The fundamental principles which are applicable to all decontamination procedure are; a. Wet decontamination method should always be used in preference to dry. b. Mild decontamination method should be tried before resorting to treatment which can damage the surfaces involved. c. Precautions must always be taken to prevent the further spread of contamination during decontamination operations. d. Where possible, contamination involving short lived activities should be isolated and segregated to allow natural decay to take its course. Decontamination of Personnel Once a radioisotope has become lodged in the body, very little can be done to increase the rate of elimination. This means that every effort must be made to prevent contamination entering the body. To this end it is vital that all personnel should obey the house rules and always wear the correct protective clothing. Even so, contamination incidents are bound to occur and so a knowledge of the current treatment is vital. The first action when dealing with a contaminated person is to ascertain whether or not he is injured. If he has a serious injury then he must be given first-aid treatment as quickly as possible. Following any necessary medical treatment, the next action are aimed at removing the contamination before decontamination can be started a careful survey must be carried out over the entire body with a suitable contamination monitor to determine the location of the contamination. In the case of partial contamination it is only necessary to contaminate the affected areas. Soap and water is the first requirement for removing contamination from the hands and other exposed areas of the skin. The soap chosen should be mild to that it will not produce skin damage after frequent use. For hands, soft-bristle nail brush should be provided for use in conjunction with soap and water over the entire surface of the hands and the wrists. Particular attention should be given to the nails, to the ridges between the fingers and to the edges of the hand. Frequent rinsing is 282 essential during the entire operation.

Radioactive Waste Disposal For the face, copious amounts of water and soap should be used, the hands alone being used to create the lather. Isolated areas of high contamination should be carefully scrubbed. All personnel should be instructed to keep the eyes and the mouth closed during treatment and to rinse the fact frequently with copious amounts of water. While using towels, or other materials suitable for drying, rubbing should be avoided. All cases of face contamination should be referred to the medical officer. Contamination of hair should be washed several times with an efficient shampoo and copious amounts of water should be used for rinsing. The latter is particularly important to ensure that contamination removed from the hair does not remain in the ears or on the face. In the event of contamination which persists even after the abovementioned procedure have been followed a number of times, the individual concerned should be referred to the medical department where more effective decontamination can be carried out under medical supervision. It is essential that skin decontamination should not be taken to the point of damaging the skin. In case of contaminated small open wounds, cuts, punctures etc., the wound should be immediately washed, bleeding should be encouraged if necessary, and the medical officer should be consulted. Whenever internal contamination occurs, it essentially becomes a medical problem, parallel in some ways to the absorption of chemical toxins. All corrective measures should be carried out under medical supervision. When contamination has been swallowed, substances designed to prevent or reduce absorption from the gastrointestinal tract, e.g. antacids or ion exchange resins, may be administered promptly after the intake. If radionuclides of high toxicity, such as Pu, are absorbed through a wound or inhaled in a soluble form. Certain chemical called chelating agents may be administered to promote excertion. Unfortunately, these substances tend to be chemically toxic themselves. The absorption of certain radioisotopes can be blocked by the prior ingestion of substantial amounts of a stable isotope of the same element. For example, the uptake of radioiodine to the thyroid can be greatly reduced by previous ingestion of a 200 mg tablet of potassium iodate. This has an important application in the even of a reactor accident. Decontamination of Working Areas A preliminary contamination survey will indicate those areas which require decontamination and such areas should be clearly marked. The decontamination measures should be restricted to those areas and every endeavor should be made to prevent the spread of contamination. The decontamination measures taken will depend upon the nature of 283 the contamination, i.e. whether it is in loose form or is reactively fixed, and

Textbook of Radiological Safety details of decontamination procedures are given in this order in the following sections. Removal of Loose Contamination Special decontamination apparatus, such as vacuum cleaners fitted with special filters, may be used to remove loose contamination. No attempt should be made to brush or dust it off, though in the case of slight contamination on the floor a wet medium such as dampened sawdust sprinkled over the contaminated area before brushing is acceptable. For all other surfaces, wet methods such as webbing are essential. The removal of contamination should be done with the minimum of rubbing and the swabs should be frequently discarded as radioactive waste. Decontamination solutions which contain complexing agents are particularly useful in such cases. Where there is copious loose contamination, a suitable strippable lacquer may be carefully applied to the contaminated surfaces. This lacquer is allowed to dry and in so doing will take up the contamination. Afterwards the strippable lacquer can be removed together with the contamination. During this operation, care should be taken in using the spraying device to avoid disturbing the loose contamination and thus giving rise to an airborne hazard. As a precaution, personnel should be in fully protective clothing. After shipping, the affected areas should be washed as described above. Removal of relatively Fixed Contamination Only wet methods should be used. The first wash should be with suitable detergent solution which will remove loose contamination and all greaseheld material. Only light rubbing should be used at this stage and the swabs should be discarded frequently as radioactive waste. Contamination remaining after this treatment should be removed by further washing with suitable decontaminating solutions. The decontaminating solutions to remain in contact with the contaminated surfaces as long as possible so that chemical reaction at the surface may assist the decontamination. Contamination remaining after several attempts at removal with the above treatment will normally be confined to small areas (unless a major spill has occurred) and further proprietary abrasive cleaners are useful for general application in this spotting treatment. Metal polish can be used to advantage on metal surfaces, abrasive creams containing a complexing agent, may be rubbed into the affected areas and left in contact for a period before being washed off. More stringent treatment would involve the use of steel wool or a similar scouring agent. If contamination persists, it will be necessary to remove the surface on which the contamination is fixed, unless it is so fixed and in such small 284 amount that it can be left in place. In the latter case, precautions should be

Radioactive Waste Disposal taken to seal in the contamination with concrete, paint or other appropriate material. Such sealed in contamination must be recorded so that in any further modifications to the building, suitable precautions can be taken against dispersing the contamination and creating a hazard. Decontamination of Equipment It is impossible to describe here the measures to be used in the decontamination of the individual pieces of equipment encountered in radiation work. However, such items of equipment may be conveniently classified into groups according to the material of which they are made, and the decontamination. Decontamination methods for equipment are of two kinds: a. Removal of contamination without damage to the surface below. b. Removal of the surface of the equipment together with the adhering contamination. In all cases the first method should be used initially and only if several attempts fail should the second method be tried, since damaged surfaces may be unsuitable for reuse because of their tendency to collect contamination easily. Decontamination of equipment should be carried out as soon as possible after its removal from the active area. Contamination left in situ over periods of time becomes fixed and becomes increasingly difficult to remove. All decontamination should be carried out using wet methods. The routine to be followed is the same for all equipment, the only difference being in the reagents used for various materials. The routine procedures are: a. Wash in detergent solution at raised temperature. This will remove all loose and grease-held contamination. This may be followed by swabbing and light scrubbing with the same solution. b. Decontaminated equipment should be washed in clean water and dried before monitoring. c. Further scrubbing and also steeping technique may be used, where contamination remains after the above treatment. In the latter, the equipment is placed in solutions of suitable decontaminating reagents, preferably at raised temperatures, and is left there for suitable periods of time. The inclusion of complexing agents in the decontaminating solution is recommended to prevent redeposition of the contamination. d. Equipment is washed in clean water on removal from the decontamination solution and is then dried before being monitored. e. Further methods will depend upon the extent and nature of residual contamination, when contamination still remains after the above treatment. If the contamination is present in spots, a treatment known as 285 spotting may be carried out using abrasive or strong acids on the small

Textbook of Radiological Safety areas involved. Where acid are used, care should be taken to ensure that the surface of the equipment is not unduly etched. If the contamination is general, it may be possible to apply abrasive over the whole surface, but if this fails, steepage in acid solution will be necessary. Precautions should be taken to prevent the acids from damaging the surface of the equipment more than is absolutely necessary to remove the contamination. Special apparatus in the form of fume hoods or gloves boxes will be necessary for the acid treatment, because of noxious fumes. f. Equipment should be well washed and dried before monitoring. Protective Clothing The protective clothing requirements in a contaminated area depend on the nature and amount of the contamination. For low levels of surface contamination an ordinary laboratory coat with overshoes and gloves may be sufficient. When there are substantial levels of airborne contamination it is usually necessary to have a fully-enclosed dry suit and a filter mask or a mark fitted with an air supply. Again, when the contamination is in liquid form, it is often necessary to wear a fully enclosed PVC suit with a filter mask or fresh air supply. Whatever the standard of protective clothing, the change and barrier arrangements must be efficient and should have the following facilities: a. Wash hand basin (and possibly a shower) and monitoring instruments (for example, a hand and clothing monitor). b. Suitable stowage on the non-active side of the barrier for the workers’ personal clothing. c. Conveniently-placed protective clothing ready for use. d. Containers for used clothing and radioactive waste. e. Notice boards at the barrier, stating ‘no unauthorized entry’, the hazards in the area, the clothing to be worn and any other precautions to be taken. f. Emergency instructions: Detailing actions in the event of possible incidents such as critically, fire, serious personal contamination, should be posted in the area. Consideration must also be given to suitable emergency exists. Special arrangements are made for laundering clothing worn in contaminated areas and the effluent from laundry facilities is treated as liquid radioactive waste. Decontamination of Protective Clothing Routine protective clothing should be cleaned regularly to avoid the buildup and fixation of contamination. For similar reasons, the clothing should 286 not be left in storage for long periods before cleaning, as experience indicates that such storage, results in the contamination being fixed and renders

Radioactive Waste Disposal decontamination increasingly difficult. Another measure taken in such clothing to avoid build-up and fixation of contamination is to provide no pockets or belts and to minimize folds. White Coats and Coveralls In small establishments it will not usually be necessary to segregate this clothing, before washing, according to the different contaminants and different levels of contamination. However, in larger establishments where a variety of radioactive material is used, it is often necessary to segregate the clothing to prevent cross contamination during the cleaning process. Contaminated clothing should be handled as little as possible, since it can give rise to airborne contamination. The process used in cleaning this type of clothing is not dissimilar to the processes adopted in conventional laundries. It is useful to remember that the more stringent washing solutions used necessitate the use of suitably resistant metals in the construction of washing machine. The actual routine of washing will depend on experience, but as a guide it is usual to give clothing two full washes (of 10 min each) with a rinse (of 5 min) in clear water after each wash. The reagents or soaps used will also depend upon experience, but a washing solution consisting of unbuild detergents, sodium metasilicate, sodium acid phosphate and citric acid has given good results with this type of clothing. The clothing should be dried thoroughly before being monitored. Rubber Gloves It is essential that the personnel wearing these gloves, and particularly the heavy rubber gloves, should wash them on completion of their work. Bulk collection of contaminated gloves is most unsatisfactory, since there is no efficient way in which they may be washed in large numbers without transferring contamination to the inside. Good-quality soap or detergents and scrubbing brushes should be provided at appropriate places where personnel may wash their gloves. The gloves should be well scrubbed and rinsed and then dried with paper towels or preferably with small pieces (say 2 x 1 ft) or toweling, which have proved economical and more satisfactory than paper towels. The small towels should be used once only and then placed in a suitable collecting bin. Respirators and Dust Masks The only satisfactory way in which these items may be cleaned is by individual swabbing with suitable detergents. The detergent used should be mild one and unlikely to cause skin complaints if it comes into contact 287 with the faces of individuals.

Textbook of Radiological Safety Small cloth or cotton swabs should be used all over and inside the facepieces. These swabs should be changed frequently. Care should be taken to clean the outside first and then the inside using clean swabs. The factpieces should be well swabbed with clean water, following the detergent treatment, and then dried. Impressive Protective Clothing On completion of an operation involving the use of this kind of clothing and where extra contamination is expected, it is essential that the operator, still wearing the suit should pass though some form of washing. This may consist of an installed shower or simply of a rinse with buckets of water. There is obviously a need to choose the correct position for this so as to prevent dispersal of contamination, and it should be followed by washing with detergent and soap solution, assisted by swabs and soft bristled brushes. Operations in these suits usually involve at least two individuals and it is a convenient practice for them to wash each other. Afterwards the suits should be rinsed in clean water and quickly dried. When dry, the suit is removed and monitored. Any residual contaminating can be treated separately with mild abrasive pastes or similar material. Suitable detergents should be used for washing these suits and it is also advantageous to use solutions containing weak citric acid and completing agents such as ethylene-diamine-tetra-acetic acid (EDTA). Footwear Rubber-soled shoes may require decontamination from time to time. The soles should be scrubbed with detergent and complexing solutions, and for resistant contamination it may be necessary to remove the surface of the rubber by the application of acetone or by mechanical buffing. Where mechanical buffing is used, it will be necessary to provide for local air extraction on the machine. The upper part of the shoes, if kept properly waxed, can be easily decontaminated. Rubber boots should be cleaned after each operation. Scrubbing in detergent and complexing solutions should be followed by the use of abrasive pastes necessary. Resistant contamination will necessitate the removal of the rubber surface by acetone or mechanical buffing. Precautions should be taken to prevent contaminated liquids from entering the boots. BIBLIOGRAPHY

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1. Chandrani L, et al. Radionuclides in Bio-Medical sciences-An introduction; Foundation books Pvt.Ltd, New Delhi 2004. 2. David JD, Hodder A, et al. The Physics of Diagnostic imaging: (2nd edn.), London 2006. 3. Kanwar Raj, et al. Management of Radioactive waste: Indian association for radiation protection, C/o RP&AD, BARC, Mumbai 2001. 4. Simon RC, et al. Physics in Nuclear medicine: (3rd edn.) Saunders, 2003.

Chapter

11

Radiation Emergency

A radiation accident is an unusual occurrence resulting from the loss of control over a radiation source which could directly or indirectly involve hazards to life, health and property. A radiation accident could occur at any stage of an operation involving radiation sources. For examples, a beam of X-ray could be inadvertently turned towards a wall to which it should not normally be directed. As a result of which persons in the adjoining room who may not even routine radiation workers could be exposed to relatively high levels of radiation. Radiation accidents would normally conform to one of the following: 1. Accidental external exposure to excessive amounts of radiation (e.g. A person remains inadvertently close to a strong source or accidentally exposed to beam of radiation). 2. Accidental spill or explosion in a working place resulting in surface and air contamination of the surroundings and contamination of personnel. In such cases the intake of radioactive substances into the body could be by inhalation, through open wounds or absorption through skin. 3. Dispersal of radioactive material to the environment as a result of an explosion, fire, mechanical shock or other incident occurring in a public place (e.g.Transport of radioactive substances). TYPE OF RADIATION ACCIDENTS 1. On-site accidents: High levels of exposure to radiation occur as a result of a person inadvertently entering a high radiation field. For example, a person walking towards the X-ray beam, when the machine is ON. In such places where there is potential for accidents, appropriate control measures should be taken well in advance to ensure that the chances of any person being accidentally exposed to high levels of radiation are minimized. Such measures includes (i) the provision of interlocks which could ensure that no person can enter the radiation area when the exposure is in progress, (ii) provision of visual or aural indication to identify high radiation level areas, (iii) provision of adequate radiation alarms which can be either located at strategic points or carried by individuals whenever they are near high radiation level areas, and (iv) the absorption of detailed administrative procedures such as provision

Textbook of Radiological Safety of suitable cordoned–off areas and prohibition of entry into such areas during radiation operations. A second category of accidental radiation exposure, which could arise in a radiation work area, could result from one of the following contingencies: i. The inability to get a remotely controlled source back into its shielded container because of mechanical or pneumatic failure. ii. Accidental breakage of a sealed source or the container of an open source, resulting in high contamination of both surface and air in the vicinity. iii. Break down of crucial ventilation systems in areas where open sources are being handled. iv. Accidents which could involve fire or explosion and which could result in the breakdown of the integrity of shielding or dispersal of radionuclides in the environment of the laboratory (e.g. criticality accident in nuclear reactor). 2. Off-site accidents: This type of accident may occur in areas to which public have access and may result from one of the following contingencies: i. Unplanned release of airborne activity to the environment of a radiation facility owing to unusual conditions such as fire, explosion, breakdown of the ventilation system or breakdown of the filter system. ii. Accident to consignments of radionuclides when such consignments are in a carrier such as a truck, train or aircraft, or when such consignments are held in storage during transit. 3. Classification of radiation accidents: Accidents involving radiation sources and radioactive minerals can be generally classified as (i) external radiation, and (ii) radioactive contamination. External radiation can result in whole body exposure,partial body exposure or localized skin exposure. Radioactive contamination may be external or internal. External contamination can occur as a result of spillage of radioactive material on skin or hands coming in contact with loose radioactive material.It can cause irradiation of the skin and underlying tissues as well as provide a potential for the material to enter the body subsequently. If the possibility is high, if the integrity of the skin is lost due to wounds, abrasions or chemicals. Internal contamination occurs most often as a result of inhalation of radioactive materials in finely divided form. As mentioned above, it may also occur when contamination present on the skin penetrates the outer layer and enters systemic circulation. Internal contamination can also occur as a result of eating with contaminated hands or consuming contaminated water; which is quite uncommon. In the case of external irradiation, first aid is not required, unless it is accompanied by traumatic injury. In radioactive contamination there is no 290 immediate risk to life. In order to minimize the long term sequela, it is

Radiation Emergency necessary to block or minimize systemic uptake and hasten the biological elimination of the contaminant. First aid, if administered to the person within a short period at the accident place is sufficient. Emergency Procedures The first step in dealing with radiation accident is to identify, segregate and treat all persons who are exposed to radiation, both external and internal. Immediate steps should be taken to assess the extent of exposure by sending the personnel monitoring TLD badges used by the exposed persons for dose evaluation. Biological monitoring and body burden measurements must also be conducted immediately. If the radiation fields are higher, special radiation measuring devices will be required. These instruments must cable of measuring much higher dose and dose rate, which are not common. Some times they are telescopically coupled to the meters, so that the detector would be in close proximity with high radiation field and the person reading the meter is away from the radiation field. Such instruments should be periodically calibrated and kept in good working conditions. Air and surface contamination samples should be analyzed urgently to take further action. The instruments required to carry out this work should also be made available. The following guidelines may be adopted during emergency. 1. Evacuate the immediate area, by ensuring that the radiation field and the extent of contamination is kept minimum. 2. Identify and isolate all persons, who might have exposed or contaminated. Arrange immediate evaluation of their TLD badges and collect samples from body fluids such as blood, urine etc. for analysis. 3. In the case of personnel contamination, carry out decontamination. 4. Regulate entry to the area of accident, so that further exposures and contamination may be prevented. 5. Notify promptly to the appropriate authorities through media such as fax, mobile and telephone, and seek suitable advice. Arrange for immediate availability of experts, who are trained to deal with emergencies. 6. Contain contamination within the accident site. In case of radioactive liquid spillage, clean up the contamination immediately. Routine protective measures such as wearing gloves and segregating the mop as radioactive waste should be adopted. If there is relatively large release of radioactive powder or aerosol in the room, that room must be immediately isolated from its surroundings by shutting off mechanical ventilation and by closing windows and doors. Entry into the room except the experts should be forbidden. A room with heavy air contamination will be decontaminated from within by drawing the air of the room through an appropriate filter.

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Textbook of Radiological Safety 7. Priority should be given to human safety and the personnel dose should be restricted with in limits (ICRP has recommended 10 rem dose limit for planned special exposures). The staff are instructed in basic emergency procedures including the persons to be contacted in case of an accident. The same may be displayed at suitable locations in the radiation installation. Mock- up operations for dealing with complicated situations associated with high radiation fields and contamination areas should always be part of a radiation emergency procedure. 8. Maintain complete records of the accident and follow up procedures. This simple instruction is often not followed, resulting in enormous complications in investigating such incidents and in the adoption of subsequent remedial measures. 9. If the accident is in the public area, the area should be cordoned off and appropriate authorities will be contacted for further action. Responsibility in the Control of Radiation Accidents

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The responsibility for controlling the use of radiation sources with in country should rest on the public authorities and the users of the source.The government should designate and define functions of those public authorities which are responsible for the control of radiation sources and dealing with radiation accidents. These public authorities will: 1. Arrange for control of the uses of radiation sources through licensing and regulations. 2. Prescribe protection standards and guidelines. 3. Establish lines of authority among national bodies. 4. Define the authority to be notified in the event of a radiation accident. 5. Establish necessary liaison with national authorities in neighboring countries. 6. Determine the need for trained personnel in the sate and arrange for training, if necessary. 7. Determine and periodically review the availability and location of trained personnel, and all other necessary equipment and services. The responsibility for immediate action following an accident originating with in the establishment will rest on the operator or user. In accordance with the requirements of his work and regulations applied by the public authorities, the user should establish an internal organization, that will: 1. Ensure that he is prepared, within the limits imposed by his resources, to deal with any accident that may occur within his premises. 2. Arrange for assistance from public authorities and other off-site organizations if necessary. 3. Provide immediate notification of the designated public authorities of accidents whose consequences may extend off-site. 4. Provide assistance to public authorities as required.

Radiation Emergency 5. Provide notification of designated public authorities of all radiation accidents. 6. Keep adequate records, and make an analysis of any radiation accidents that occur. Emergency Protective Clothing Provision of the protective clothing is usually indicated in areas where the operations concerned will expose the personnel involved to a high risk of contamination and of breathing contaminated air. Before wearing emergency protective clothing, including fully impervious clothing, one has to have a full change, which implies removal of all personnel clothing and the wearing of simple clothing supplied by the laboratory. A typical change would be drill trousers underwear, shirt, socks and shoes. Fully Impervious Clothing This consists of garment so designed that they cover the individual completely, except for the head and neck and the hand and feet with a layer of impervious material. As a breathing apparatus to be worn the same time, hoods of similar material are used. Finally complete exclusion from contamination is obtained by wearing rubber gauntlet gloves pulled well over the cuffs of the protective suit and rubber boots with the bottom of the impervious suit trousers brought over them. Pressurized Clothing This is suit made of impervious material which completely encloses the individual. Such a suit effectively isolates the individual inside from any contamination on the surfaces or in the air. Compressed air supplied to the suit enables normal breathing during operations. The compressed air line delivers its air immediately in front of the face, as this arrangement provides plenty of air for breathing and at the same time helps to reduce the misting of the transparent head piece. Complete protection of the hands and wrists is afforded by wearing rubber gloves, which are securely taped to the suit to prevent the ingress of any contamination. Rubber boots are usually worn with the suit. Assistance is always necessary to dress an individual in any impervious clothing. Such assistance is particularly essential when impervious clothing is being removed. If the individual concerned undresses it is more than likely that he will become contaminated from the active material present on that suit. Breathing Apparatus When compressed air supply is not available for use in the pressurized suits described above, breathing sets may be used. These comprise a well fitted face piece in which suitable goggles are inserted. This face piece is 293

Textbook of Radiological Safety attached to a cylinder. Regulating valves attached to the cylinders enable the wearer to control his air supply. This breathing set may be used in conjunction with the fully protective impervious clothing and, so desired, the individual concerned can enter a contaminated atmosphere under emergency or maintenance conditions. DIAGNOSTIC RADIOLOGY-SKIN INJURIES

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In India eight incidents have been reported in medical applications. Six of these incidents were associated with teletherapy units, due to which 24 persons received low radiation doses and one person received a high dose on hand, requiring amputation. Two incidents were reported with x-ray fluoroscopy, during cardiac catheterization and pace maker insertion. The two patients received very high doses to the skin of their back. Interventional radiology using fluoroscopy equipment for image guidance may pose greater risk of X-ray induced injuries. Interventional radiology finds application in Cardiology, General radiology, and Neuroradiology. The reason for higher risk includes (i) more extended periods of time, (ii) multiple use of radiography, (iii) relatively high radiation exposure for both patients and personnel. About 700,000 procedures are carried out / year globally and more than 70 injuries have been reported. Fluoroscopy exposure may leads to deterministic effects, that includes erythema, skin ulcer, temporary epilation, and dermal fibrosis. But these effects vary with fluoroscopy time, dose rate, fractionation of dose, mode of operation, patient age, and site of exposure. Skin injuries have been reported in patients undergone prolonged fluoroscopy guided interventional procedures such as percutaneous transluminal coronary angioplasty (PTCA), radiofrequency cardiac catheter ablation, vascular embolization etc. The cumulative skin doses in some patients may exceed 10Gy, which is sufficient to cause radiation induced skin injury (Fig. 11.1 to 11.6).These injuries may appear only after a threshold period of several months. The risk of deterministic and stochastic effects of radiation exposure varies for different areas of the skin. The age of the patient, undergoing long fluoroscopic time of imaging is also an important factor. The most common interventional procedures like PTCA and percutaneous translumunal angioplasty (PTA) are used in patient population over 40 years of age. Small numbers of adults and children are also treated using interventional procedures. It is essential to differentiate between interventional procedures in adults, in young adults and in children. If areas of the skin are likely to be exposed to levels of absorbed doses that approach the skin tolerance, the patient should be informed in advance about the possible effects of treatment. All relevant parameters should be documented for all interventional procedures.

Radiation Emergency

Figs 11.1A to C: A 49 year old woman with 8 year history of refractory supraventricular tachycardia A–C, Photographs show sharply demarcated erythema above right elbow at 3 weeks after radiofrequency cardiac catheter ablation (A), tissue necrosis 5 months after procedure (B), and deep ulceration with exposure of the humerus at 6.5 months (C), (AJR:177, July 2001) (For color version see plate 5)

Fig. 11.2: 56 year old man with obstructing lesion of right coronary artery. Photograph of right postero-lateral chest wall at 10 weeks after percutaneous transluminal coronary angioplasty shows 12 × 6.5 cm hyperpigmented plaque with hyperkeratosis below right axilla (For color version see plate 5)

Fig. 11.3: A 75 year old woman with 90% stenosis of right coronary artery. Photograph of right lateral chest obtained 10 months after percutaneous transluminal coronary angioplasty shows area of hyper- and hypopigmentation, skin atrophy, and telangiectasia (poikiloderma) (For color version see plate 6)

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Fig. 11.4: Skin necrosis (erythema and hyper-pigmentation), 3 years after angioplasty examination (For color version see plate 6)

Fig. 11.5: A 17 year old girl with history of cardiac arrhythmia underwent two cardiac ablation procedures in 13 months. Photograph taken 2 years after last intervention shows atrophic indurated plaque with skin telangiectasia at right lateral chest wall involving posterolateral aspect of breast. Induration resulted in limited movement of right arm. Risk of breast cancer is increased (For color version see plate 7)

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Fig.11.6: A 49 year old man with history of liver cirrhosis and intractable upper gastrointestinal bleeding who underwent two transjugular intrahepatic portosystemic shunt (TIPS) placements and one attempted TIPS placement within a week. Photographs show progression of ulceration. (A) Secondary ulceration with surrounding rings of de- and hyperpigmentation 6 months later. (B) Small blisters developed at 7.5 months after procedure. Wound is very painful. (C) Wound has progressed in size and depth at 10 months. (D) Nonhealing ulcer with exposure of deep tissues, including spinous process of vertebra, at 22 months. (E) At 23 months, musculocutaneous skin grafting was performed. Disfigurement is permanent.(AJR:177, July 2001) (For color version see plate 7)

Radiation Emergency NUCLEAR MEDICINE: RADIATION ACCIDENTS The IAEA report 17 (2000) has reported 7 accidents in the case of unsealed sources. A therapeutic dose of 370 MBq was prescribed to a wrong patient because two patients had the same name. Among the two patients, one for iodine administration and other for treatment of a lung disease, with identical name were there. The physician familiar with the patient was not available and another physician was assigned to administer the dose. The lung disease patient was administered with I-31 of 370 MBq activity. In another event a patient came for a diagnostic bone scan was administered with 333 MBq of I-131,instead of 740 MBq of Tc-99m for bone scan study. In another event a 60 year old women was referred to the nuclear medicine for thyroid ablation. The physician prescribed 6475 MBq of I-131 by oral administration. The technician with out reading the labels picked up two vials containing 6475 MBq and 5180 MBq of I-131 and administered both the activity to the patient. In another event a patient was prescribed 370 MBq for a thyroid treatment. A capsule containing 370 MBq was ordered. The supplier sent a capsule containing 444 MBq. Though the technician calibrated the dose, but misread the activity of 444 as 370 MBq. The patient was administered with 444 MBq, which resulted in an overdosage of 20%. Another event includes, administration of 180 MBq of I-131 for a whole body scan, to a feeding mother. The scan indicated an unusual high breast uptake of I-13l, which resulted a 300 Gy thyroid dose and 0.17 Gy whole body dose to the infant. Both the physician and technologists have failed to confirm that the patient was not breast feeding. In another event a therapy dose of 7400 MBq was administered to an 87 year old patient. After 34 hours the patient experienced cardiac failure and 16 staff members attempted to resuscitate the patient, to insert pacemaker etc. Blood and urine contaminated with radioactivity were spilled, but the personnel clothing were not checked for activity. A dose of 0.3 Gy was reported in one of the nursing staff. In another event a patient was to be administered with 259 MBq of I131.There were two capsule of 130 MBq each in the vial, which were labeled correctly. When the vial is inverted only one capsule fell out and the same was administered. That is patient was given 130 MBq instead of 259 MBq acitivity, 50% of the prescribed dose. Radiation Emergencies in Radioisotope Laboratory Radiation emergencies in radioisotope laboratory would normally involve only spillage of radioactive liquids. The prepardness and procedures to meet such emergencies are discussed below:

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Textbook of Radiological Safety Emergency Preparedness 1. Charts detailing various steps to be taken by the radiation workers in case of emergency should be conspicuously displayed in the laboratory. 2. All radiation monitoring and measuring instruments should be routinely checked and kept in working condition.Handling equipment such as tongs, forceps etc. must be kept in ready access. 3. The ventilation system of the radioisotope laboratory should be routinely checked and maintained properly. 4. Ready availability of a decontamination kit containing all the items of decontamination should be ensured to deal with an accidental spill effectively. 5. A proper inventory of radioisotopes received, used and disposed should be maintained. Emergency Procedures It is very difficult to make rules to manage variety of radiation accidents. However, spillage of radioactivity is the most likely accident in radioisotope laboratory, which require the following procedures to be recommended in dealing with such emergencies: 1. Confine the spill immediately, by droping paper towels or other absorbent material into it. 2. Evacuate the immediate area so that persons will not walk over the spill and spread the contamination. 3. If the spilled material has splashed on to a person or clothing, immediate steps should be taken to remove the laboratory coats or outer garments and to leave them in the contaminated area. Hands and other contaminated areas on the body should be washed thoroughly with soap and water. Care should be taken not to abrade or inflame the skin surface. If internal contamination has taken place, immediate action should be taken to minimize the deposit of radioactivity in internal organs and tissue and enhance the excretion of the ingested radioactive material, under expert medical supervision. Bioassay or whole body counting, if facilities are available, should be carried out to confirm internal contamination. 4. Contaminated area should be decontaminated by experienced persons wearing surgical gloves, shoe covers and a surgical face mask if available. Tongs or forceps should be used to remove the contamination by absorbent material. The absorbent material so collected should be kept in a polythene bag to be treated as radioactive waste. After as much contamination as possible has been removed in this way, the surface should be washed with damp-not wet-paper towels held by forceps, always working towards the centre of contaminated area, rather than away from it. 298

Radiation Emergency 5. A contamination monitor should be used to monitor the area as well as personnel during the procedure of decontamination. The contamination monitor should be operated by someone who is not involved in the clean up, so that the instrument does not become contaminated when the decontamination procedure is over. The contaminated gloves, shoe covers etc. should be kept in a polythene bag for decay or ultimate disposal as radioactive waste. The forceps / tongs should be kept separately covered in polythene bag for decay of radioactivity in it. 6. In case of a large release of radioactive powder or aerosol in a room, such a room must be immediately isolated from its surroundings by shutting off mechanical ventilation and by closing windows and doors. A room with heavy air contamination can be decontaminated from with in by drawing air of the room through an appropriate filler. 7. Maintain complete records of the accident giving details of the radioisotope, activity involved etc. and follow up procedures. For effective management of any kind of radiation emergency, the RSO should educate and familiarize all radiation workers with the steps to be taken to meet the emergency. Use of radionuclides in research applications will not give significant exposure to general public, provided certain basic precautions are taken. These precautions mainly involve good accuracy for the radionuclides, good work practice and a well controlled programme for the disposal of radioactive waste. Procedures for Handling Spills Accidental spills of radioactive materials are quite common in nuclear medicine departments, but they are not life threatening hazards. Hence spills should be treated as events completely with out hazard and the staff should aware of the procedures to be followed. The steps involved in the radioactive spill are: – To inform, – To contain, and – To decontaminate. 1. Individuals in the immediate work area should be informed that the spill has occurred so that they can avoid contamination if possible. Individuals outside the immediate area should be warned so that they do not enter it. The RSO should be informed so that he /she may begin supervising further action as soon as possible. 2. The laboratory personnel should attempt to control the spill to prevent further spread of contamination, without risking themselves. A flask that has been tipped over should be uprighted. Absorbent pads should be thrown over a liquid spill. Doors should be closed to prevent the escape of airborne radioactivity. The spill area should be closed off to 299

Textbook of Radiological Safety prevent entry, especially by persons who might not be aware of the spill. Personnel monitoring for contamination should be started as soon as possible, so that contamination and uncontaminated persons can be segregated. To prevent further spread of radioactivity, contaminated individuals should not be allowed to leave the area until they are decontaminated. The uncontaminated individuals should not be allowed to enter the spill area. Contamination monitoring should be done using a sensitive radiation monitoring instrument appropriate for the type of radioactivity involved. It is advisable that each laboratory have on hand a thin window GM counter survey meter for handling such situations. 3. Personnel decontamination procedures should receive first priority, followed by decontamination of work areas, etc. Personnel involved in decontamination procedures should wear protective clothing to avoid becoming contaminated themselves in the process. Contaminated skin should be flushed thoroughly with water. Special attention should be given to open wounds and contamination around the eyes, nose and mouth. Contaminated clothing should be removed and placed in plastic bags for storage. After major localized areas of personnel contamination have been attended to, a shower bath may be required to remove more widely distributed contamination. Decontamination of laboratory and work areas should not be attempted except under the supervision of RSO. If the work surfaces and floors are constructed from a non absorbent material, soap and water is generally used for decontamination. Contaminated areas should be cleaned “from outside in” to minimize the spread of contamination. Porous or cracked surfaces may create difficult problems. If complete decontamination is not possible, it may be necessary to cover and shield the affected surfaces or perhaps even to remove and replace them. RADIOTHERAPY: RADIATION EMERGENCIES A situation in a beam therapy facility can be considered as an emergency, if it could result in higher actual or potential radiation doses than those in normal situations to personnel, patients or public. In 2000, the International atomic energy Agency (IAEA) published its report No 17 “Lessons learned from accidental exposures in Radiotherapy”. It described 92 accidents resulting in an incorrect dose to the patient. Out of 92, 32 of these accidents were related to the use of sealed sources (brachytherapy). In the case of telegamma units, two potentially hazardous situations may arise: i. The source does not, or only partially, go back to OFF position; in other words, it continues to be in the “ON” position even after the termination of the set treatment. 300 ii. The source falls off from the source head.

Radiation Emergency The type of situations where the telegamma source continues to be in the “ON” position, vary with the source “ON-OFF” mechanism in the unit. They include failure of (i) pneumatic system (ii) shutter to close, and (iii) drum to rotate, to bring the source back to “OFF” position at the end of the set treatment time. The emergency situation continues until the source is completely shielded by mechanical manual or other means. In certain systems emergency may caused by leakage of mercury which functions as a shield, thus leaving the source in the “ON” position even after the set treatment time is over. In medical accelerators, normally such type of emergency does not occur. However, recently a few cases of computer software mix-up, resulting in the wrong selection of electron mode, have been reported. The following paragraph discusses several instance of such emergency situations in teletherapy units which have occurred in this country and abroad. Emergency Situations Leakage of Mercury (Shutter) In September 1980, the film badges worn by the radiation therapist, the physicist and the technologist of a hospital recorded high doses. On investigation it was noted that this was caused by an emergency situation of an Eldorado A Unit 2. Here the source is stationary and shielded in “OFF” position by mercury between the source and the collimator. When the source is made “ON” the mercury is pumped out and stored in a reservoir and held there against gravity. Small mass of mercury started leaking from this reservoir when the source was “ON”. The leakage of mercury was noted and they arranged to collect the leaking mercury and put it back into the unit as immediate remedial measure. But some quantity of mercury was lost in this process and hence the efficiency of mercury as the shield was gradually reduced. This process continued until December 23, 1980, when the treatment by the unit had to be discontinued for want of spare parts. Later, the hospital staff developed a special filter, which when incorporated into the unit, completely prevented the leakage of mercury. A number of overexposures to personnel were reported during this period. Based on chromosome aberration test reports, the individual equivalent dose, ranging from 20 to 50 rem were estimated to have been caused by this emergency. Failure of Shutter Movement 1. In July 1984 a case of malfunctioning of the mobile shutter of a TheratronB unit was reported. The shutter mechanism was working only when the unit was pointed downwards. The mobile shutters were encountering the obstructions on its path and as a result, the shutter used to remain partially open during the “OFF” position. The source was unloaded and the shutter were removed and examined. It was observed that the mobile 301

Textbook of Radiological Safety shutter had lost its integrity and that a tiny particle removed from it got embedded in the housing and caused the obstruction. The shutter was thoroughly cleaned, polished and lubricated and the unit re commissioned. During this emergency the doses to personnel involved in the repair work were negligible. 2. The shutter mechanism of a Gammatron unit, serviced during the source replacement on December 13, 1984, failed to function and remained open at the end of set treatment time on January 23, 1985. The service engineers reported that the shutter developed extraordinary friction. Treatment was stopped until April 16, 1985, when the source was unloaded and shutter repaired. The individual doses received by personnel during the repair were less than 10 mrem. 3. In another emergency on November 23, 1985 the shutter of a Gammatron unit remained in open position and could not be closed, even manually. The service engineers reported that the source had to be unloaded prior to shutter repair. It was ascertained that the shutter mechanism had developed extraordinary friction due to damage in the ball bearing movement system. For want of spare parts, to be imported, the unit could not be used for treatment for about 16 months. The source was unloaded on April14, 1987 and the unit was repaired. The personnel dose during the repair was less than 10 mrem. 4. In another case of Gammatron unit, the shutter failed to close at the end of the set treatment time in October, 1986. The patient was immediately removed and defect was rectified by the hospital engineer. Pneumatic System Failure Recently, two cases of source “stuck” in the “ON” position for GammarexR unit have been reported. In both cases, the source drawer did not return to the parking position at the end of the set treatment time, and the sources were pushed back to “OFF” position manually with the T-rod. Personnel exposure were well within permissible levels. In one case, subsequently the source was unloaded, the drawer and its passage area thoroughly cleaned and the source replaced. Improper Functioning of Timer

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A typical case of improper functioning of the timer of Picker unit has been reported. The timer was observed to reset the time automatically at the end of the set treatment time. This resulted in the source to be in the “ON” position continuously until the timer was manually interrupted. Even though such an incident has not been reported in India, a formal intimation regarding this malfunction was sent to all users of Picker unit. It was confirmed that the timers in the Picker units in India were functioning properly.

Radiation Emergency Software Mix-up in Accelerator Two incidents of massive overdoses of patients took place in a hospital in USA from Therac-25 accelerator on March 21 and April 11, 1986, with “Malfunction-54” occurring in both the cases. A third case with the same accelerator which occurred on April 25, 1985 was reported later. Another similar incident from a different hospital using the same model of accelerator occurred on January 17, 1987, though the reported malfunction was of a different nature. The first two patients died and the third patient was seriously maimed. The “Malfunction-54” occurred when the technician inadvertently selected photon beam mode for a patient who was to be treated with electron beam of lower energy. The operator realizing the mistake promptly corrected for the mode using the edit-key facility. The computer displayed the electron beam mode and the “beam ON” command was given to start the treatment. The treatment was abruptly terminated after one second with the patient writhing in pain calling for help from inside treatment room. The physicist later simulated the malfunction and measured the dose in water phantom as 25000 rads in one second. AECL subsequently informed all Therac-25 users to suspend the use of this accelerator. A plan for correcting the errant software was submitted by AECL to Food and Drugs Administration, U.S.A. In the meantime the fourth incident occurred where although the photon beam mode was selected the heavy metal target did not come into position into the path of the electron beam and instead a light field mirror came in position to intercept the electron beam. Corrective action plan was amended accordingly. San Jose, Costa Rica A radiotherapy overexposure occurred at the San Juan de Dios Hospital in San jose, Costa Rica in august and September 1996. About 115 patients who were treated for neoplasm by radiotherapy were affected. Out of the 115 patients 42 died and 73 were alive. The radiation exposure was a major factor for death in 3 patients and partial in 4 patients. The event occurred on August 22,1996,when a Cobalt-60 source was replaced. After the source loading, the new source was calibrated, and an error was made in calculating the dose rate. This miscalculation resulted administration of higher doses to the patients than prescribed. Panama Accident A radiation accident occurred in the National institute of oncology, Panama in May 2001,in which 28 patients undergoing radiotherapy in cobalt unit were affected. The IAEA team visited the place on May 22, 2001 and reported that out of 28,eight patients had died. The death of five is due to radiation 303

Textbook of Radiological Safety over exposure, one death is related to cancer and remaining two death, no conclusion was made. The cause for the accident due to wrong data entry into the treatment planning system (TPS) computer. The spatial coordinates of the shielding block has to be fed into the TPS in way that one shield at time. From August 2000, the coordinates of the all the shielding blocks were entered as a single block, which resulted incorrect dose and treatment times. Lack of written procedures and manual check when the data input procedure was changed, resulted radiation overexposure to the above patients. Ottawa Hospital Cancer Center A relocation of an orthovoltage treatment unit from one campus to the other in the Ottawa hospital cancer centre in the Fall 2004, was followed by an error at the time of recommissioning. The error was reflected as incorrect output tables for all 4 beam qualities and for all field sizes other than 10 x 10 cm2. A total of 1019 patient treatments were delivered to 620 patients using the incorrect orthovoltage output tables during November 2004 to November 2007. Patients were under dosed to a maximum of 17 %. No overexposure was reported and the cause was due to the omission of a back scatter conversion factor for all the fields other than 10 x 10 cm2 (www.iaea.org). Overexposure During Source Exchange Two radiation workers, while working on a source exchange of a decayed cobalt-60 source in a hospital in the state of Sao Paulo, Brazil on july 21, 2008, received significant overexposure. The workers transferred the source drawer from the unit head to the transfer cask with the help of tool. When the tool was pulled out, the source drawer momentarily came out of the transfer flask before the worker pushed back. They received a whole body equivalent dose of 135 mSv and 0.7 mSv respectively. Later on August 15, 2008 it was found that one of the fingers of the individual was blistering. The extremity dose was estimated as 25 Sv. Different types and degrees of emergency as discussed above can potentially occur in beam therapy facilities. The facility and the radiation safety officer must be prepared to meet such kinds of emergencies at any time. Line of action to be followed may vary with the type of emergency and the personnel must be clearly aware of the same Radiation Emergencies-Fall of Source Drawer During source replacement or source head repair of telegamma therapy units with source “ON-OFF” mechanism, serious accidents can take place unless proper care is taken. Recently, two such serious accidents 304 have taken place in India in two different institutions. In each case, sources

Radiation Emergency loaded in parallopiped shape source drawer fell down from the source head during repairing of shutter mechanism. In one case, it was cobale-60 source of 6000 RHM and in the other it was Caecium-137 source of 2050 Ci. 1. First Incident: The first incident took place when two under trainedstaff members of the institution tried to repair the shutter mechanism of a Gammatron-3 teletherapy unit. They removed the gear box mounted at the back side of the drum shutter to look for any foreign particles obstructing the movement. Later they removed the front plate as well as the arresting strip of the drawer. The head was tilted through +300 in the forward direction and the unit was rotated to bring it to a convenient position. The source drawer, which was lying freely in the cavity slipped down from the head. Realizing the seriousness of the situation, they came out and locked the room immediately. They received whole body equivalent dose of 29.50 mSv (2950 mrem) and 12.40 mSv (1240 mrem). Source Retrieval Operation The retrieval operation was planned on the basis of information collected during a preliminary visit to assess the situation. The details available from the layout of the installation were also useful in planning the management of the accident. Temporary extra shielding was planned for secondary walls to be provided by piling up sand bags upto 1 meter height. An aluminium tray with a groove like depression at the bottom to sit on the source drawer (Fig. 11.7) was designed and fabricated. It was planned to cover the source drawer with the above tray and to fill it up with lead shots through a polystyrene pipe by personnel positioned behind the maze wall.

Fig. 11.7: Aluminium tray designed to cover the source drawer

Area monitoring of the entrance to the treatment area and other areas of interest were carried out using a Teleflex cable of 12 m length 305

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connected to the detector, keeping the meter of the monitoring instrument outside the room. The radiation levels at various points measured before and after providing temporary shielding are given in table 11.1. The various location in which radiations were measured are given in Fig 11.8. Since there was no lead glass viewing window and since the CCTV was focused to the couch, a mirror mounted on a wheel trolley was employed at the entrance of the treatment room to locate the source drawer. With the help of this mirror, it was observed that the major length of the source drawer was hidden by the base drum of the couch, which was locked to the ground. Further, a raised wooden platform made of interlocked portions was provided in the room for the convenience of patients as it lowered the height of the couch. The source drawer was lying on this platform. As such, positioning the aluminium tray to cover the source drawer and to fill it up with lead shorts were not viable to persons positioned behind the temporary shielding erected at the entrance position. A rehearsal of the operation with a dummy drawer was conducted in the adjacent teletherapy room. It was concluded that aluminum tray can be placed over the source drawer by one person entering the room with a remote handling tong of 2.0 m length and that the equivalent dose for this operation will be about 15 mSv (1500 mrem). Actually, this job could be carried out with an equivalent dose of 12.5 mSv (1250 mrem) to the operator. In the wall common to control room and treatment room, a tapered hole was drilled at the height of 2.5 m from the floor for the passage of polystyrene pipe to facilitate filling up of the aluminum tray had to be shielded, temporarily. Lead shots (about 40 Kg) packed in cloth bags were used for this purpose. This was carried out for a collective equivalent dose of 0.025 mansievert (2.5 manrem). The exposure rate at the drilling position was then reduced to 5 mR/h from 200 mR /h. Polystyrene pipes of 5 cm in diameter and 6 meter in length were extended to the aluminium tray through the hole drilled on the wall at Control room. About 350 Kg of lead shorts of diameter 1-2 mm were poured into the aluminium tray through this pipe. This brought down the radiation level inside the treatment room to a workable level. All unwanted materials and a major portion of the wooden platform inside the room were quickly removed. The source drawer was lying at the middle the platform. To transfer the drawer into a transport flask, the drawer hade to be brought to the edge of the platform. Another aluminium tray of slightly larger size and smaller grove for source drawer to exactly fit in (to reduce streaming) was fabricated and positioned in line with the aluminum tray with lead shots in it, which was covering the source drawer. The source drawer was then transferred to the groove of the second tray. After emptying

Radiation Emergency the first tray, it was used in line with the second tray for shifting the source drawer further close to the edge of the platform to enable it to be loaded into the transportation flask. Before loading source drawer into the transportation flask, shielded source drawer had to be lifted through 20 cm to bring the source drawer in line with the cavity loading the fallen source drawer into transportation flask, the teletherapy unit was repaired and subsequently the source loading was carried out. The equivalent doses received by personnel involved in the retrieval operation are given in Table 11.1. The total collective equivalent dose was about 0.07 man sievert (7.0 man rem). Table 11.1: Radiation levels in and around the Telecobalt facility before and after providing temporary shielding at the entrance and maze region (The source drawer with 60Co source of 222 TBq (6,000 Ci) was lying on the wooden platform inside the treatment room. The source was facing the ceiling. The locations where radiation levels were measured are numbered and marked in Fig.11.8) Radiation levels in and around the Telecobalt facility Before providing temporary shielding

After providing temporary shielding

Location Location number description

Air kerma Exposure rate (mGy h-1 ) rate (mR h-1)

Air kerma rate (mGy h-1)

Exposure rate,(mR h-1)

1.

1.74

200

0.35

40

0.87 0.44 0.87 1.74 6.97 43.50 522.00 104.40 0.17 0.02

100 50 100 200 800 5 × 103 60 × 103 12 × 103 20 2

0.87 0.44 0.87 1.74 1.31 5.22 522.00 43.50 0.04 8.70 × 103

100 50 100 200 150 600 60 × 103 5 × 103 4 1

Behind the maze wall

2. 3. 4 5. 6. 7. 8. 9. 10. 11.

Entrance to treat room Door Control panel

2. Second incident: In this case a source drawer containing 2050 Ci Caesium137 fell down during repairing of the shutter mechanism of a Ceasa Gammatron unit by the service engineers, who missed to arrest the source drawer. Retrieval operation The radiation levels in the control room and surrounding areas were not very high and operation of retrieval could be managed from the entrance door. The fallen source drawer was covered by an aluminium tray and it was filled with lead shots. A second compact tray of mild steel mounted on a mild steel plate with three hooks on each side was fabricated and 307 placed in line with the first aluminium tray. The second tray was filled

Textbook of Radiological Safety

Fig. 11.8: Lay out of the teletherapy facility, where radiation levels are measured

with lead bricks and lead shots and source drawer was transferred below this tray (on the M.S. Plate). By using chain pulley system the source drawer shielded under the mild steel tray and resting on the mild steel plate was lifted and placed on a trolley which was provided with 10 cm thick layers of interlocking lead bricks covering the entire base of the mild steel tray. Additional shielding by interlocking lead bricks were provided on three sides and on the top of the tray and the trolley was positioned in a corner of the room. Later the source drawer was loaded into the teletherapy head after the repair of the unit. The total collective equivalent dose in the operation was about 0.02 mansievert (2 manrem). From the above incidents we can conclude that unless proper care is taken during servicing or source loading, serious accidents involving high exposure to personnel and long down time of machine can take place. In addition, skillful planning of management of accident is necessary for keeping radiation dose to the minimum to personnel involved in the operation. BRACHYTHERAPY: RADIATION ACCIDENTS The IAEA report No 17, presents 32 accidents related to Brachytherapy .The type and number of accidents is summarized in the Table 11.2. Errors in the specification of the source activity, dose calculation or the quantities and units resulted in doses that were up to 170% of the prescribed dose. 308 Some accidents were related to human mistakes, for example, the use of an incorrect source due to fading of the color coding. This is listed under

Radiation Emergency “other” in the table, which also includes accidents caused by badly implanted sources, removal of the sources by the patient or otherwise dislodged sources. The most severe accident was due to equipment failure, where a lethal dose was delivered to a patient. Table 11.2: Type and number of accidents reported in Brachytherapy treatments (IAEA 2000) Accident caused by

Number of cases

Dose calculation error Error in quantities and units Incorrect source strength Equipment failure Other Total

6 2 7 4 13 32

HDR Brachytherapy units have higher potential for damage to patients as a result of misadministration. In one HDR event the patient was prescribed a dose of 35 Gy to lung. There was a kink in the catheter which positioned the source 26 cm away from the target site. The error was not discovered until the treatment was over. Because of this error, hypopharynx received 35 Gy and the target received only 1 Gy. The other events includes using a HDR prescription for the wrong patient, and malfunction of the HDR equipment etc. Over all the treatment delivery errors can be summarized as follows: positioning of the active source train outside the treatment volume, mispositioning of dose prescription points and failure to detect a source that separated from its cable and remained in the implanted catheter for 91 hours. IAEA reported that the misadministration is due to malfunctioning equipment, such as catheters do not allow full movement of sources. It also mentioned the inadequacy of personnel training which caused the misadministration. The above demonstrates the need for a well designed programme of quality assurance in any Brachytherapy department. Its goal should be the consistency of the administration of each individual treatment, the realization of the clinical intent of the radiation oncologist, and the safe execution of the treatment with regard to the patient and staff. EMERGENCY PREPAREDNESS: ACTIONS In spite of preventive measures, accidents or un usual events may occur. The safety assessment will identify accident scenarios and situations, and counter measures for mitigation are designed. Some of the actions in the emergency response plan need to be taken immediately, with out hesitation or mistakes. To achieve this personnel need to be trained and simulation drills carried out. A clear and concise list of actions and responsible officers 309 is posted in relevant places. There is a written documentation of rules for

Textbook of Radiological Safety action and training, including simulating exercises, for the emergency response plan, which needs to be periodically reviewed. Line of Command of Actions The RSO must establish the line of command of actions to be taken, in the case of any emergency. This will include the type of immediate action to be taken and persons to be contacted. The addresses and telephone numbers, if any, of these persons must be conspicuously displayed. Source not Returning to “OFF” Position Removal of the Patient If the source does not return to “OFF” position after the set treatment time is over, the technician must remove the patient without himselft getting directly exposed to the primary beam. This would also imply that an audiotype gamma zone monitor is available, calibrated and working, which will indicate that the source is “ON” or “OFF”. To facilitate such action, the RSO must routinely advise the workers regarding the route to be taken by them inside the treatment room for various directions of primary beam. It must be clearly demonstrated to the staff that the dose equivalent to the worker for such a procedure will be only of the order of 10-20 mrem. They must also be told to be in the treatment room only for the minimum time required and, in any case, not to expose any part of their body including hands, to the primary beam. After the patient is removed, the technician must lock the room, inform the RSO immediately, and wait for further instruction. Restoring the Source to “OFF” Position The RSO must try to return the source to the “OFF” position as specified for the unit concerned and with appropriate care. This may involve procedure such as pushing the source drawer to the parking position with a T-rod, pressing a popped-up button on the source head, rotating a wheel in the gantry, or by switching off electric power as specifically directed in the Instruction Manual supplied by the manufacturer. If this procedure does not result in returning the source to the “OFF” position, the RSO must leave the treatment room. The room must be closed immediately and the service engineers of the company concerned may be summoned for appropriate assistance. Administrative action will include vacating for shielding adjacent rooms and areas and cordoning off the facility as required. Requirement of Equipments and Accessories The follow up materials and equipments must be available in the hospital

310 and be kept outside the telegamma room. It must be possible to procure

Radiation Emergency items such as sand bags, lead shots, etc. at very short notice in case of emergency. 1. Survey meter: Properly calibrated and maintained survey meters of appropriate ranges must be available at all ranges. A wide range survey meter with long cable facility is desirable. 2. T-rod: T-rod or other such equipment meant for the unit. 3. Manual: Operation/service manual of the unit which will give details regarding type of source drawer etc. must be kept at easy accesses. In one of the recent major incidents the instruction manual was located only a after prolonged search. 4. Binoculars: Binoculars could be of help to verify whether the source is in the “ON” or “OFF” position, as also its location, if it has fallen down. 5. Lead shots: 300-400 kilograms of small sized lead shots, of about 1-2 mm diameter. This will be needed to shield the source in case it falls off from the unit. 6. Lead wool: Several kilograms of lead wool preferably packed in bags will be needed to be thrown over the source from a safe distance, so that the source can be initially shielded, before its actual retrieval. This will enable personnel to approach closer to the source for the retrieval work. 7. Lead bricks: Several lead bricks of standard size, regular and interlocking type, will be useful to shield the source further, once the lead shots and lead wool are thrown to cover the source. 8. Sand bags: Generally, a telegamma-particularly telecobalt-room is designed such that two walls will act as primary walls and the remaining two as secondary walls. However, if the source falls on the floor, all walls become primary walls. Hence, additional shielding may be needed for the secondary walls, and in some cases, depending on the location in the room where the source has fallen, for primary wall too. A number of sand bags may need to be arranged on these walls, so that the exposure rates outside are within permissible levels. 9. Long handled tongs: Long handled tongs on which the radiation probes could be fixed will be needed to monitor the radiation levels at various locations in the telegamma room. This will help in deciding the course of action to be followed. 10. Long pipe and funnel: A long pipe is needed so that the lead shots can be dropped through a funnel connected to one end kept near the door or maze in such a way that radiation is avoided by the worker to the extent possible. The pipe must be in an inclined position, in such a way that the other end is over the source. 11. Transport container: The transport container (source flask) must be available locally with the servicing firm, so that remedial action can be initiated immediately in the case of the source fall. It must also be 311

Textbook of Radiological Safety stressed that in case of major repairs of the head, drawer or shutter, the source, if needed, must be unloaded prior to such work. 12. Personnel monitoring dosimeters: All personnel in the telegamma room including casual workers must wear personnel monitoring dosimeters. Legal and Regulatory Requirements Rule 20 of Atomic Energy (Radiation protection) rules, 2004 stipulates that the employer is the custodian of radiation sources in his possession and shall ensure physical security of the sources at all times. The employer shall inform the competent authority, within 24 hours, of any accident involving a source or loss of source of which he is a custodian. The worker shall inform the licensee and RSO, of any accident or potentially hazardous situation that may come to his notice. The licensee shall prepare emergency response plans and submit to the competent authority for approval. It is the duty of the radiological safety officer to investigate and initiate prompt and suitable remedial measures in respect of any situation including emergencies that could lead to radiation hazards. He should develop suitable emergency response plans to deal with accidents and maintaining emergency preparedness. The RSO must send a report of the emergency to the Head, Radiological safety division, AERB, MUMBAI. The report should give details of the incident, including the causes, remedial action, dose received and steps taken to avoid such emergencies in future. Prevention of Emergency Generally, emergencies arise in the case of old and poorly maintained units. It must be stressed that proper work discipline, as well as regular servicing, maintenance and quality assurance tests of the teletherapy unit and associated equipment will definitely help to prevent potential emergencies. It must be pointed out that most of the major incidents involving telegamma units have occurred during source transfer or repair work. It must be ensured that any servicing of the source head must be done only by experienced engineers and in the presence of the RSO. In many cases, emergencies arise because of temptations to compromise on or by-pass of, simple requirements of radiation safety. These should never be resorted to. Typical example include arrestor not provided to prevent movement of source drawer and transport container not immobilized during source transfer operation. Periodic drills must be arranged by the RSO to simulate emergency situations. This will also help to avoid tendency for complacency among the staff.

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Radiation Emergency MEDICAL MANAGEMENT OF PERSONNEL EXPOSED TO RADIATION Partial Body Exposures – Localized Exposure (Radiation Burns) In localized exposure, a small part of the body is exposed to radiation. Most commonly, hands, feet, legs and face are irradiated but any other part of the body may also be involved. Skin is the first organ to be exposed following irradiation. The earliest damage seen in transient erythema which comes immediately after exposure and is due to dilation of capillaries resulting from histamine like substances released by injured cells. This is followed 2-4 weeks later by fixed erythema which may come in waves is much deeper and more prolonged than the transient erythema. If the dose is more than 3Gy (300 rad) epilation, dry, moist desquamation and ultimately necrosis of the epidermis result. Long term sequelae are pigmentation, atrophy of dermis, sweat glands, sebaceous glands and hair follicles, fibrosis of dermis and increased susceptibility to trauma and chronic ulceration. Damage to the germinal cells in the basal layer is critical in the pathogenesis of erythema and desquamation. It is the dose to these cells that determines the severity of skin damage. Skin Effects Following Exposure Transient Erythema Early transient erythema may occur in a matter of hours following doses of more than 2 Gy, because of changes in permeability of capillaries. The main wave of erythema peaks at 10 days to 2 weeks and requires a larger dose of about 6 Gy. Epilation Epilation or hair loss, occurs if there is sufficient reduction in the replicative capacity of germinal cells or the matrix of the hair follicles. Temporary epilation may occur after doses of about 3 Gy, with an on set at about 3 weeks and regrowth requiring 5 weeks or more. Epilation is permanent if the dose exceeds about 7 Gy. Some of the radiation effects and the levels at which these may occur are given below in Table 11.3. Dry Desquamation Dryoeneum, much like a sun burn, may occur after single doses of more than 14 Gy, because of depopulation of clonogenic cells in the epidermis. Healing requires the repopulation of basal cells from surviving clonogens.

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Textbook of Radiological Safety Table 11.3: Radiation effects and radiation dose Radiation effects

Threshold dose of X and γ rays

Time of onset

2 Gy 3 Gy 14 Gy 24 Gy 15 Gy 18 Gy 10 Gy

2-24 h 3 wk 4 wk 6 wk 8-10 wk >10 wk >52 wk

Early transient erythema Temporary epilation Dry desquamation Secondary Ulceration Late erythema Ischemic dermal necrosis Telangiectasis

Moist Desquamation Moist desquamation requires higher doses greater than 18 Gy and also results from depopulation of clonogenic cells in the epidermis. Healing is caused by repopulation of surviving clonogens or micration of clonogenes from the edges of the irradiated area. These effects may cause substantial discomfort, but provided they are not severe, they heal and clear up as the population of basal cells recovers. Transepidermal Burn This is similar to second degree thermal burns with a latent period of 1-2 weeks. Radiation burns are sometimes deceptive on superficial appearance as damage to important organs in subcutaneous tissue nerve endings, hair follicles, sweat glands, endothelium of blood vessels may not be obvious. Among these, the injury to the endothelium of blood vessels is the most serious. It produces endartritis obliterans, leading to necrosis of overlying tissues, which continues to progress for several months. The severity of burns depends on the dose and dose-rate and doses of 30 Gy (300 rad) or above blistering and skin loss may take place. In such cases, besides subcutaneous tissues other internal structures are affected and may give rise to radiation necrosis of bone, muscle and other internal organs. Initial symptoms are erythema, pain, swelling, itching, or tingling and epilation. Full Thickness Radiation Burn This is similar to third degree burns and a serious version of transepidermal injury. The injury extends up to the dermis and produces prompt and severe pain. In case damage to circulation is present, the healing will take long time and surgical intervention may be required. Pain is an important feature of the exposure of skin, particularly in the extremities, to high doses of radiation. This pain is maximum with the appearance of vascular lesions. The pain is experienced during the first few days, lasting several hours and it may last for long periods.

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Radiation Emergency Long Term Effects Long-term sequelae, include dermal atropy, telangiectasia, and necrosis. These effects occur months to years after higher doses of radiation (10-18 Gy) and are caused by primarily by vascular damage to the dermis. Late effects also may develop after unusually severe early effects, in which case they are referred to as consequential late effects. Management of Radiation Burns History A detailed history of accident with name, age and sex of the person, the nature of radiation and energy, possibility of whole body exposure or contamination etc. should be collected. Personal TLD badges will provide some idea about the exposure. Some times the patient may not aware of irradiation and dose. Complete examination of the skin repeatedly on the first day is required to see is there any prodromal erythema. The time at which transient erythema occurred, along with other symptoms, will enable the physician to come to a rough conclusion regarding the dose and the ultimate prognosis, with the development of fixed erythema. Investigation The following investigations and procedures are recommended: 1. Complete blood count 2. Blood lymphocyte culture and Chromosomal analysis 3. Sperm count 4. Culture and antibiotic sensitivity test 5. Estimation of radionuclides in urine and stools 6. Serial color photography 7. Thermography 8. Non invasive vascular studies 9. Radioisotope scintigraphy 10. Slit lamp examination of eyes 11. Physical dosimetry Samples should be taken immediately for items 1-5 in the above list. Concurrently photographs should be taken and dosimeters sent for evaluation of dose. Scintigraphy may be done before slit examination in view of blood contamination with 99mTC. Even after the area of burn becomes apparent, the underlying damage cannot be observed with accuracy clinically. Thermography and scitigraphy offer a means of detecting areas affected significantly by localized irradiation, and the functional status of the organ. This information is helpful in planning any surgical intervention with out waiting for the clinical symptoms to unfold fully.

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Textbook of Radiological Safety If there is leucopenia in the first week, it is suggestive of whole body exposure. There is danger of infection which should be treated vigorously. Surgical intervention should be kept to the minimum during this phase of bone marrow depression which usually lasts about 4-8 weeks. Specific Treatment 1. Mild erythema: This may become dry and start itching in 3-4 weeks. A bland lotion or steroid ointment should be applied locally. No tight clothing should be worn on the affected part. 2. Transepidermal burn: Pain should be relieved by analgesics, and drug like phenylbutazone,which cause bone marrow depression should be used. Sterile protective dressings should be used. Systemic antibiotics should be given for prevention of infection. Usually the burns will heal without skin grafting in the absence of infection. 3. Full thickness radiation burn: The burns may progress from initial blistering to skin loss and deep tissue necrosis, giving rise to severe pain,tissue loss and infections. This will require surgical intervention, the timing of which will be difficult to decide due to slow progression of burn. Bone marrow depression may further complicate the condition. In case there is leucopenia at 2-6 weeks, surgical treatment should be kept at minimum until haematopoietic recovery takes place (usually in about 6-8 weeks). In case the involved area is more than a few square cms (2-3 sq.cm) skin grafting will be required. Larger areas involving necrosis and gangrene of distal portions of fingers of extremities will require amputation. In beta-ray burns, early excision and skin grafting may spare the patient from pain and discomfort. Lastly, follow up of such cases is important because healed radiation burns may result in weak atrophic skin that is subject to chronic and recurrent ulceration. The time for amputation and reconstructive surgery depends on the following determinants: i. Intractable pain ii. Size and location of injury iii. Degree of control over infection iv. Degree to which vascular damage can be estimated v. Value of the part. BIBLIOGRAPHY 1. A practical guide to quality control of Brachytherapy equipment: ESTRO Booklet No.8 2004. 2. AERB safety code: Medical management of persons exposed in Radiation accidents.AERB/SG/MED-1, 1990. 3. Chandrani L, et al. Radionuclides in Bio-Medical sciences-An introduction; Foundation books Pvt. Ltd, New Delhi 2004.

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Radiation Emergency 4. Jerrold TB, Seiber, JA, Edwin ML, John MB. The essential physics of medical Imaging, (2nd edn.) Lippincott Williams and Wilkins 2002. 5. Lecture notes: Training course on Radiation safety for Radiation therapy technologists; RSD, AERB and RPAD, BARC, Mumbai. 6. Safety report series No 17: Lessons learned from accidental exposures in radiotherapy, IAEA,Vienna, 2000.

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Index A Absolute risk 26 Absorbed dose-rad/gray 2 Accuracy of corrections for count losses 143 Actions and precautions that can reduce radiation exposure to the fetus 234 Additional installation requirements 87 Additional requirements for type A packages 250 Additional requirements for type B packages 251 AERB classification of radioisotope laboratories 188 AERB guidelines for starting radioisotope laboratory 187 AERB guidelines to set up a radioimmunoassay (RIA) laboratory 189 AERB specification for layout of radiotherapy facility 194 Air conditioning 85 Alignment of table gantry 134 Annual limit on intake 8 Applicator integrity 160 Aprons 280 Area monitoring 103 Area survey 113, 115 Artificial sources 10 Assistance to patients 206 Associated facility 87 Atomic energy act-1962 167 Atomic energy regulatory board 167 Avoid of pregnancy after radionuclide therapy 237 Avoid of pregnancy after receiving radiotherapy for breast cancer treatment 242

B BEIR report V and VII risk estimate 26 Biologic effects 17 Booking, storage, transport and delivery of package 257 Brachytherapy facility design 91 Brachytherapy sources, equipment and installations 192

Brachytherapy: radiation accidents 308 Breastfeeding 233 Breathing apparatus 281, 293 Burial of solid waste 273

C Calibration and maintenance of radiation monitoring instruments 117 Carbon fiber materials 210 Cardiac catheterization and pregnancy 223 Category III-yellow 249 Category II-yellow 249 Category I-white 249 Ceiling mounted barriers 205 Ceiling 55 Cell 14 Central beam alignment 122 Chance of approaching dose limits of exposure 226 Chemical purity 150 Chemical treatment 271 Chest and extremity radiography in pregnancy 222 Classification of waste 269 Cobalt-teletherapy machine survey 114 Collective dose 6 Collective effective dose equivalent 7 Collimator axis, light beam axis and cross-hairs coincidence 155 Collimator rotation 155 Collimator test 135 Committed dose 6 Computed tomography installation 70 Concentrate and contain 268 Conduit 83 Congruence of radiation and optical fields 121 Consent 184 Consentee 185 Consignor's declaration 258 Construction materials 84 Consumer products 11 Contamination control 229 Continuation of work of a pregnant employee in X-ray department 224

Textbook of Radiological Safety Control of PH 151 Control of starting materials 148 Controlled and uncontrolled areas 67 Cosmic rays 9 Counseling of patients 226 Counting rate performance 139 CT and pregnancy 222 CT number linearity 147

D Decontamination of equipment 285 Decontamination of personnel 282 Decontamination of protective clothing 286 Decontamination of working areas 283 Decontamination procedures 281 Delay and decay 268 Determination of particle size 151 Deterministic effect 20 Detriment 8 Diagnostic radiology-skin injuries 294 Dilute and disperse 268 Disposal of low activity wastes into the environment 276 Disposal of P-32 and I-131 into municipal sewers by medical users 277 Disposal of radioactive effluent into the ground 276 Disposal of radioactive solid waste 272 Disposal of radioactive waste from nuclear medicine procedures 278 Distance 32, 41 Doors and interlocks 80 Dose limits to patients 199 Dose limits 197 Dose philosophy 197 Dose reduction in fluoroscopy 213 Dose reduction methods in paediatric chest CT 220 Dose reduction methods in pediatric abdominal CT 221 Dose reduction methods in pediatric radiography 217 Dose reduction methods 219 Dry desquamation 313 Ducts and shielding 82

E

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Early somatic effects (partial body irradiation) 19

Early somatic effects (whole body irradiation) 18 Effective dose or effective dose equivalent 4 Effects on radiation exposure in utero (ICRP-84) 221 Electrical and mechanical tests 145, 159 Electron beam data 158 Emergency preparedness: actions 298, 309 Emergency procedures 291, 298 Emergency protective clothing 293 Emergency situations 301 Employer 183 Energy resolution 139 Energy 158 Enhanced natural sources 10 Epilation 313 Equipment and peripherals 207, 216 Equipment 74 Equivalent dose 4 Establishing a diagnostic X-ray facility 65 Establishing a nuclear medicine facility 71 Establishing a radiotherapy facility 79 Evaporation 271 Excepted packages 247 Exposure - roentgen 2 Exposure rate constant 3

F Facility design and construction 78 Facility design for brachytherapy 59 Facility design for diagnostic X-rays 42 Facility design for nuclear medicine 47 Facility design for radiotherapy 48 Failure of shutter movement 301 Fetus risk 28 Field area 207 Field flatness 158 Field symmetry 158 Film badge 96 Filtration 207 Fissile packages 248 Flatness and symmetry 159 Fluoroscopy installation 68 Focal spot size 122 Focus to table top distance 134 Footwear 280, 288 Free drop test 250

Index Full thickness radiation burn 314 Fully impervious clothing 293

G Gantry and couch 146 Gantry rotation 155 Gantry tilt 135 Gaseous waste 275 General checks 108 General guidelines 64, 71 General precautions 240 General radiography installation 68 General requirements 249 Genetic effects 21 Genetic risk 27 Genetically significant dose 7 GM type survey meters 106 Guidelines for using TLD badge 99

H Half value layer 35 HDR brachytherapy survey 116 HDR treatment rooms 93 Head leakage source on position 114 Head leakage-source off condition 114 Hereditary effects 19 High contrast resolution 136 High contrast sensitivity 134

I IDR and TADR 49 Image intensifier assembly leakage 134 Image intensifier 211 Image quality tests 146 Image quality-attenuation and scatter correction accuracy 144 Image receptors 210 Image uniformity and pixel noise 146 Imaging rooms 74 Impressive protective clothing 288 Improper functioning of timer 302 Incineration 272 Industrial packages 248 Information to carriers 260 Inspection of the equipment 108 Instantaneous dose rate method 52 Instruments and accessories 109 Interaction of radiation with tissue 14 Internal radionuclides 10 Intrinsic resolution 138

Investigation 315 In-vitro and radioimmunoassay (RIA) 75 In-vivo diagnostic facility 74 Ion exchange 271 Ionization chamber survey meter 104

J Jaw symmetry 154

K Kerma 2

L Labeling and identity 231 Labeling of the package 254 Laboratory coat 280 Late somatic effects 19 LDR and MDR treatment rooms 92 Lead apron 204 Leakage and contamination 163 Leakage limits for brachytherapy 40 Leakage limits for cobalt teletherapy 39 Leakage limits for X-ray housing 39 Leakage of mercury (shutter) 301 Leakage radiation 38 Legal and regulatory requirements 312 Licensing 78 Line of command of actions 310 Linear accelerator 113 Linear energy transfer 16 Linearity of MA station 127 Linearity of timer 128 Liquid waste 270 Long term effects 315 Low and high contrast resolution 147 Low contrast resolution 136 Low contrast sensitivity 134

M Mammography installation 69 Management of cadavers containing radionuclides 187 Management of radiation burns 315 Marking of package 253 Maternal hydration 237 Measurement of computed tomography dose index (CTDI) 136 Measurement of MA linearity 136

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Textbook of Radiological Safety Measurement of operating potential 135 Measurement of timer linearity 136 Mechanical isocenter 155 Mechanical test 251, 134 Medical exposure 11 Medical management of personnel exposed to radiation 313 Minimum facilities required 189 Model plan 88 Moist desquamation 314 Multiple beam alignment check 157 MTC-99 generators 280

N Natural radiation source 9 Nuclear fuel cycle 11 Nuclear medicine physician 186 Nuclear medicine technologist 186 Nuclear medicine: radiation accidents 297

O Occupancy factor (t) 41 Occupancy in the room 206 Occupational exposure 11 Ongoing evaluation of product performance 152 Optical and radiation beam congruence 155 Optically stimulated luminance dosimeter 100 Organ shield 205 Ottawa hospital cancer center 304 Output consistency 128, 136 Overalls (boiler suit) 280 Overexposure during source exchange 304 Overshoes 281

P

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Package handling 252 Package radiation levels 253 Packaging and package requirements 249 Panama accident 303 Partial body exposures - localized exposure (radiation burns) 313 Particulate contamination 151 Patient dose reduction 229

Patient motion 211, 217 Patient observation and communication 81 Patient positioning 211 Patient waste 280 Pediatric computed tomography 218 Pediatric exposure 229 Pediatric fluoroscopy 218 Pediatric radiography 216 Penetration test 251 Performance evaluation of CT 145 Periodic QA schedule 164 Personnel monitoring systems and features 102 Personnel requirements 182, 185, 193 Personnel responsibilities 182, 185, 193 Personnel safety during source transfer operations of teletherapy and HDR brachytherapy units 238 Personnel wear 228 Photon beam data 158 Pin prick method 161 Placards 256 Pneumatic system failure 302 Pocket dosimeter 100 Positional accuracy 160 Pregnancy and radiation protection in nuclear medicine 233 Pregnancy and radiation protection in radiotherapy 240 Pregnancy and radiation 221 Pregnant patient with cervical carcinoma 242 Pregnant staff and continuation of work 237 Pregnant woman and radiotherapy 241 Preparation of package for transport 253 Pressurized clothing 293 Prevention of emergency 312 Primary barrier adequacy 115 Primary barriers 50 Primary radiation 37 Procedure for authorization 190 Procedures for handling spills 299 Protection in computed tomography 214 Protection in fluoroscopy 212 Protection in nuclear imaging 227 Protection in pediatric imaging 215

Index Protection in radiography 206 Protection in radionuclide therapy 232 Protective barrier design 40 Protective clothing 286 Protective devices 204, 227

Q QA for fluoroscopy X-ray unit 134 QA for gamma camera 138 QA for HDR brachytherapy 159 QA for linear accelerator 154 QA for mammography X-ray unit 133 QA for radiopharmaceuticals 147 QA for single photon emission computed tomography (SPECT) 140 QA for treatment planning system 164 Quality assurance for computed tomography 134 Quality assurance for diagnostic radiology 119 Quality assurance for nuclear medicine 137 Quality assurance for PET-CT 141 Quality assurance for radiography units 120 Quality assurance for radiotherapy 154 Quality assurance test format 129

R Radiation dose from patients 228 Radiation dose tests 136 Radiation effects in utero 23 Radiation effects on DNA 22 Radiation emergencies in radioisotope laboratory 297 Radiation emergencies-fall of source drawer 304 Radiation exposure level (XT or P) 41 Radiation induced cancer 20 Radiation isocenter 156 Radiation limits for shielding design 199 Radiation profile width 135 Radiation protection rules-2004 168 Radiation risk 24 Radiation safety 147 Radiation sensitivity and profile widths 146 Radiation survey in diagnostic radiology 107

Radiation survey in nuclear medicine 111 Radiation survey in radiotherapy 112 Radiation survey 107, 154 Radiation units 2 Radioactive fallout 11 Radiochemical purity 149 Radiography 204 Radioiodine therapy and pregnancy 235 Radiological safety officer (RSO) 183, 186 Radiologist 183 Radionuclide activity 148 Radionuclide purity 148 Radionuclide therapy 78, 111 Radiopharmacy 77 Radiotherapy: radiation emergencies 300 Reduction of fetal dose when a pregnant patient undergoes radiotherapy 241 Regulatory controls for diagnostic X-ray equipment and installations 181 Regulatory controls for nuclear medicine facilities 184 Removal of loose contamination 284 Removal of relatively fixed contamination 284 Removal of the patient 310 Requirement of equipments and accessories 310 Respirators and dust masks 287 Responsibility in the control of radiation accidents 292 Restoring the source to "off" position 310 Reverse osmosis 271 RHM and RMM 3 Risk models 25 Risk to a pregnant woman, when a family member is treated with radioiodine 236 RMM and curie 3 Room lighting and lasers 83 Routine protective clothing 280 Rubber boots 281 Rubber gloves 280, 287 Rule 1. Short title, extent and commencement 168

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Rule 10. Suspension, modification or withdrawal of a license 172 Rule 11. Modification of radiation installation or change in working condition 172 Rule 12. Restrictions on use of sources 172 Rule 13. Restriction on certain practices 172 Rule 14. Radiation symbol or warning sign 173 Rule 15. Dose limits and other regulatory constraints 173 Rule 16. Safety standards and safety codes 173 Rule 17. Prohibition of employment of persons below certain age 173 Rule 18. Classified worker 173 Rule 19. Radiological safety officer 173 Rule 2. Definitions 169 Rule 20. Responsibilities of the employer 174 Rule 21. Responsibilities of the licensee 174 Rule 22. Responsibilities of the radiological safety officer 175 Rule 23. Responsibilities of worker 176 Rule 24. Records of workers 177 Rule 25. Health surveillance of workers 177 Rule 26. Medical exposures 177 Rule 27. Radiation surveillance requirements 178 Rule 28. Directives in the cases of exposures in excess of regulatory constraints 178 Rule 29. Power to appoint or recognize persons or agencies 178 Rule 3. Licence 169 Rule 30. Inspection of premises, radiation installations and conveyances 179 Rule 31. Power to investigate, seal or seize radiation installation or radioactive material and to give direction to the employer 180 Rule 32. Directives in case of accidents 180 Rule 33. Emergency preparedness 180

Rule 34. Decommissioning of radiation installation 181 Rule 35. Offences and penalties 181 Rule 4. Fees for license 170 Rule 5. Exemption 170 Rule 6. Exclusion 170 Rule 7. Conditions precedent to the issuance of a license 171 Rule 8. Issuance of license 172 Rule 9. Period of validity of license 172

S Safety work practices that can reduce internal radiation dose 230 San Jose, Costa Rica 303 Scan localization light accuracy 135 Scatter fraction and count rates 142 Scattered radiation 38 Sea dumping 274 Secondary barrier for leakage radiation 55 Secondary barrier for scattered radiation, when the primary beam strikes the wall 53, 54 Service engineer 183 Shielding calculation 47 Shielding calculation for diagnostic X-ray 42 Shielding design for computed tomography 46 Shielding 34 Skin and surface contamination 281 Skin effects following exposure 313 Software mix-up in accelerator 303 Solid waste 272 Somatic risk 27 Source not returning to "off" position 310 Source on position leakage 113 Source strength verification 163 Source to object distance 209 Sources and nature of waste 268 Sources of exposure 37 Sources of radiation 9 Spatial linearity 138 Spatial resolution 141 Specific treatment 316 Spillage 112 Stacking test 251 Staff protection 237

Index Staggered autoradiography 161 Steps to be taken in patients found to be pregnant after administration of radioiodine therapy 236 Sterility and apyrogenicity 152 Stochastic effect 20 Storage in transit requirements 252 Storage 274 Student/trainee 183 Survey procedure 109 Swipe test 111 System resolution 138 System sensitivity 140

T Table position /increment 135 Table top exposure rate 134 Teletherapy installation 190 Temporal accuracy 162 Ten day rule and its present status 29 Tenth value layer 36 Termination of pregnancy after radiation exposure 223 Terrestrial radiations 9 Test for type A package 250 Test for type B package 251 Thermal test 252 Thermoluminescent dosimeter 97 Thyroid shield and lead glass 205 Timer checking 125 Timer error 162 Timer linearity 162 Total filtration 126 Training 79 Transepidermal burn 314 Transient erythema 313 Transport index 248

Treatment control area 81 Treatment of pregnant patients with radionuclides 235 Tremcard 259 Tube housing leakage 128, 137 Tube voltage (KVP) 124, 207 Type A packages 246 Type B packages 247 Type of radiation accidents 289 Type-I (simple) 200 Type-II (medium) 201 Type-III (stringent) 201 Types of packages 246 Types of radioactive waste 270 Types of RIA laboratories 76

U Uniformity 138 Use factor (U) 41

W Warning signs and lights 86 Waste management 267 Water immersion test 252 Water spray test 250 Weekly dose rate method 50 White coats and coveralls 287 Width of the primary wall 52 Woman of childbearing age and nuclear medicine examinations 233 Workload (W) 40, 42, 46, 109 Work practice 206 Workers 198

X X-ray generator 135, 146 X-ray technologist 183

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