Addre Ad dress ssin ing g th the e In Incr crea easi sing ng Re Rele leva vanc nce e of of Ir Irra radi diat atio ion n Assisted Stress Corrosion Cracking R.Gérard Internat Inte rnationa ionall seminar seminar « Netwo Networking rking for for effectiv effective e R&D », Petten, Petten, 22-23 22-23 Septem September ber 2003
Relevance of IASCC
Known problems:
PWR: baffle bolt cracking BWR: core shroud cracking (IGSCC due to thermal sensitization with possible assistance of radiation effect?)
Potential problems:
PWR: baffle bolts/plate swelling
PWR: long term evolution of IASCC susceptibility
Maximum neutron dose (approx.) - 40 years of operation
PWR (typical 900 MW unit)
Core barrel
Core baffle
Baffle bolt
≈
≈
≈
10 dpa, T 300 – 330°C ≈
80 to 100 dpa T 370°C (at former levels) ≈
70 dpa, T
≈
350°C
BWR (typical)
Shroud and top guide
Fuel support
≈
7 dpa
≈
0.6 dpa
PWR lower internals (Typical)
PWR core barrel
Baffle-formers assembly
Baffle/formers/core barrel assembly (detail)
CW 316 SS 5/8” bolts With or without cooling holes
Baffle bolt cracking
Cracked baffle bolts detected in French, Belgian, Japanese and U.S. plants (316 CW and 347 stainless steels) Examination of extracted bolts confirm IASCC origin of the cracking Number of bolts largely exceeds what is necessary to maintain the integrity of the internals in accident conditions (control rods insertion) no safety issue as long as cracking remains limited No reliable predictive model to date; influence of dose and stress
Baffle bolt cracking evolution in French CP0 units and Tihange 1 100
Fessenheim 1/Tihange 1 *
* The bolts of the plants represented by the same symbol are made from the same heat
Fessenheim 2 / Bugey 2 *
90
Bugey3 S T L O B D E K C A R C D E M R I F N O C F O R E B M U N
Bugey4
80
Bugey5 70
Bugey 2
60
50
Fessenheim 2
40
30
Tihange 1
20
10
Bugey 3 Fessenheim 1
Bugey 5 Bugey 4
0 50000
65000
80000
95000
110000
125000
140000
EQUIVALENT FULL POWER HOURS
From “An analysis of Bafle/Former bolt cracking in Franch PWR” Peter M. Scott, M.C.Meunier, D.Deydier, S.Sylvestre and A.Trenty, Environmentally Assisted Cracking : Predictive methods for Risk Assessment and Evaluation of Materials , ASTM 1401 “
Irradiation swelling
Problem known from fast reactors experience
Until recently, not expected to occur at PWR temperatures
First measurement on non-negligible swelling reported in Tihange 1 Baffle Bolt in 2000 (CIR): 0.25% max Lower levels measured in baffle bolts extracted from U.S.plants (0.03%) Irradiation swelling is highly temperature dependant
Irradiation swelling
Likely to affect only limited locations in PWRs (high dose and temperature) like “re-entrant corners” baffle/former junctions No indication to date of significant risk for PWR internals in short/medium term Uncertainty on long term (very high dose) behaviour
BWR core shroud cracking
BWR - Core shroud cracking
First case in 1990: Mühleberg, Switzerland
More than 30 reactors affected worldwide
Materials: AISI 304, 304L, 347, 316L
IGSCC/IASCC initiated in HAZ
International activities
Cooperative IASCC research program (CIR-II)
International IASCC advisory committee
EPRI MRP (Materials Reliability Program) Internals Task Group
JoBB (Joint Baffle Bolt Program)
Japanese projects (JAPEIC)
OECD Halden Reactor Project
European projects
INTERWELD (FP5): irradiation effect on the evolution of microstructure, properties and residual stresses in the HAZ of stainless steel welds PRIS (FP5) : properties of irradiated stainless steels for predicting lifetime of NPP components
LIRES (FP 5): development of reference electrode
AMALIA network
PERFECT proposal (FP 6)
What do we need?
Reasonable understanding of IASCC and irradiation swelling in order to develop predictive models. need for experimental data (microstructure, microchemistry, mechanical properties, dose, temperature) on service components (LWR) irradiated to high dose.
Mitigation methods. In spite of (tens of) millions of EUR spent in the last years in various international or national programs, we are still far from reaching these objectives!
What are the problems?
Any test on highly irradiated stainless steel is very expensive. Lot of information was obtained on specimens irradiated in the fast reactor BOR 60; not certain it is representative of PWR irradiations at lower fluxes (no formal proof of the contrary either). Lack of material from real components irradiated to very high doses. Limited flow of information between groups (proprietary data) nobody has the full picture.
Problems
Increasingly difficult to get money from the industry to finance very costly international programs which do not deliver solutions.
Mitigation/repair techniques
“Engineering” solutions Baffle bolt cracking: Inspection + replacement (in most cases same material with slightly improved design) BWR core shroud cracking:
repair (tie-rods, clamps, brackets) or replacement (performed in Japan: Fukushima 2 and 3, Sweden: Oskarshamn 1, Forsmark 1-2). Difficulty and outage time very variable depending on design
Other possible mitigating actions
Improved material (need improved resistance to IASCC and swelling; hard to qualify because would need long irradiation in representative conditions; accelerated irradiations not necessarily representative)
Improved design (stress concentration, temperature)
Improved chemistry (HWC in BWRs, Noble metal addition)
Conclusions
Progress was made in the last years on many aspects of IASCC (microstructural and microchemistry evolution). Understanding of the phenomenon still limited in spite of ambitious (and expensive) international programs. In-plant problems under control for the short/medium term thanks to “engineering” solutions (inspections, replacement, repair). Questions remain concerning PWR internal degradation at very high doses (40 years or more).
