Reactor pressure vessel annealing -- Effective mitigation method
- Nuclear Research Inst. Rez plc (Czech Republic)
- SKODA Nuclear Machinery Ltd., Plzen (Czech Republic)
Reactor pressure vessels of old generation were mostly manufactured from materials with high content of impurities which results in high increase in irradiation embrittlement values. Standard mitigation methods for decrease this damage--application of low-leakage core or dummy elements insertion--are inefficient if applied during the reactor operation. Thermal annealing of reactor pressure vessels has been shown as a very effective method for restoration of initial material properties in a high extent. Even though annealing process is not fully understood from the microstructural changes point of view, results from the testing were so promising that many annealing of WWER RPVs were performed. Nevertheless, some problems still remains, connected mainly with monitoring of the extent of annealing restoration as well as with re-embrittlement rate after such a properties restoration. Experience with WWER-440 RPVs is discussed, mainly because of the austenitic cladding existence. Cladding does not allow to take templates from the inner RPV surface and it is damaged during operation, as well. At the same time, no monitoring of cladding behavior during operation was planned within surveillance programs. Problems connected with material behavior monitoring after annealing as well as during further operation (re-embrittlement rate) are discussed together with the assessment of inaccuracies and possible solutions.
- OSTI ID:
- 277826
- Report Number(s):
- CONF-960306-; ISBN 0-7918-1226-X; TRN: 96:018235
- Resource Relation:
- Conference: ICONE 4: ASME/JSME international conference on nuclear engineering, New Orleans, LA (United States), 10-13 Mar 1996; Other Information: PBD: 1996; Related Information: Is Part Of ICONE-4: Proceedings. Volume 5: Radioactive waste disposal; Decontamination and decommissioning; Aging assessment and license renewals; Global advances in nuclear codes and standards; Major component reliability; Rao, A.S. [ed.] [General Electric Nuclear Energy, San Jose, CA (United States)]; Duffey, R.B. [ed.] [Brookhaven National Lab., Upton, NY (United States)]; Elias, D. [ed.] [Commonwealth Edison, Downers Grove, IL (United States)]; PB: 525 p.
- Country of Publication:
- United States
- Language:
- English
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