(Reactor vessel embrittlement)
Technical Report
·
OSTI ID:6757878
The scope of the exchange meeting included US and USSR presentations regarding (1) annealing of reactor vessels (the USSR has annealed nine vessels thus far), (2) research on reactor-vessel materials from reactors taken out of service (Shippingport Gundremmingen, and Novovoronezh), (3) radiation embrittlement of materials for the vessel in VVER-1000 type reactors, (4) application of embrittlement data to operating reactors, (5) mechanisms of radiation damage to vessel materials, and (6) pressurized-thermal-shock issues, including a comparison of US and USSR calculated values of the conditional probability of vessel failure. The traveler was primarily concerned with item 6. The USSR revealed that they evaluate vessel integrity using a deterministic approach for which the design-basis transients are defined with the aid of event trees. These transients are less severe than the most severe transient considered in the US Integrated Pressurized-Thermal-Shock (IPTS) analysis, but the USSR believes their approach is more conservative because the most severe US transients have a very low anticipated frequency of occurrence. The USSR is considering a probabilistic approach to the evaluation of vessel integrity and have compared their results with ours. Differences exist and the reasons for the differences were identified.
- Research Organization:
- Oak Ridge National Lab., TN (USA)
- Sponsoring Organization:
- NRC
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 6757878
- Report Number(s):
- ORNL/FTR-3662; ON: DE90014301
- Country of Publication:
- United States
- Language:
- English
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FAILURES
HEAT TREATMENTS
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PWR TYPE REACTORS
RADIATION EFFECTS
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