Addressing the Increasing Relevance of Irradiation Assisted Stress Corrosion Cracking

R.Gérard

International seminar « Networking for effective R&D », Petten, 22-23 September

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 ≈ 10 dpa, T≈ 300 – 330°C

Core baffle ≈ 80 to 100 dpa T≈ 370°C (at former levels)

Baffle bolt ≈ 70 dpa, T ≈ 350°C

BWR (typical)Shroud and top guide ≈ 0.6 dpa

Fuel support ≈ 7 dpa

PWR lower internals (Typical)

PWR core barrel

Baffle-formers assembly

Baffle/formers/core barrel assembly (detail)

CW 316 SS 5/8” boltsWith 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 crackingNumber 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 limitedNo reliable predictive model to date; influence of dose and stress

Baffle bolt cracking evolution in French CP0 units and Tihange 1

Fessenheim 1

Fessenheim 2

Bugey 3

Bugey 4Bugey 5

Bugey 2

Tihange 1

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50000 65000 80000 95000 110000 125000 140000

EQUIVALENT FULL POWER HOURS

NU

MB

ER O

F C

ON

FIR

MED

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AC

KED

BO

LTS

Fessenheim 1/Tihange 1 *Fessenheim 2 / Bugey 2 *Bugey3Bugey4Bugey5

* The bolts of the plants represented by the same

symbol are made from the same heat

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 experienceUntil recently, not expected to occur at PWR temperaturesFirst measurement on non-negligible swelling reported in Tihange 1 Baffle Bolt in 2000 (CIR): 0.25% maxLower levels measured in baffle bolts extracted from U.S.plants (0.03%)Irradiation swelling is highly temperature dependant

Irradiation swelling

No indication to date of significant risk for PWR internals in short/medium termUncertainty on long term (very high dose) behaviour

Likely to affect only limited locations in PWRs (high dose and temperature) like “re-entrant corners” baffle/former junctions

BWR core shroudcracking

BWR - Core shroud cracking

First case in 1990: Mühleberg, SwitzerlandMore than 30 reactors affected worldwideMaterials: AISI 304, 304L, 347, 316LIGSCC/IASCC initiated in HAZ

International activities

Cooperative IASCC research program (CIR-II)International IASCC advisory committeeEPRI MRP (Materials Reliability Program) Internals Task GroupJoBB (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 weldsPRIS (FP5) : properties of irradiated stainless steels for predicting lifetime of NPP componentsLIRES (FP 5): development of reference electrodeAMALIA networkPERFECT 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 invarious 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” solutionsBaffle 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).