Application of the failure assessment diagram to the evaluation of pressure-temperature limits for a pressurized water reactor

Conference · Fri Jun 01 04:00:00 EDT 1984 · Am. Soc. Mech. Eng., (Pap.); (United States)

OSTI ID:6291248

Yoon, K K; Bloom, J M; Pavinich, W A; Slager, H W

The failure assessment diagram approach, an elastic-plastic fracture mechanics procedure based on the J-integral concept, was used in the evaluation of pressure-temperature (P-T) limits for the beltline region of the vessel of a pressurized water reactor. The main objective of this paper is to illustrate the application of an alternate fracture mechanics method for the evaluation of pressure-temperature limits, as allowed by Title 10, Code of Federal Regulation Part 50 (10 CFR 50), Appendix G. The evaluation of P-T limits for the beltline region of a pressurized water reactor vessel was based on the following assumptions: ASME Pressure Vessel and Piping Code, Section III, Appendix G reference flaw End-of-life fluence level in the beltline region Longitudinal flaw in the beltline weld J-resistance material toughness curves obtained from the U.S. Nuclear Regulatory Commission's Heavy Section Steel Technology (HSST) program Other material properties obtained from the Babcock and Wilcox Integrated Reactor Vessel Material Surveillance Program The maximum allowable pressure levels were calculated at 33 time points along the given bulk coolant temperature history representing the normal operation of a pressurized water reactor. The results of the calculations showed that adequate margins of safety on operating pressure for the critical weld in the beltline of the pressurized water reactor vessel are assured.

• Research Organization: Babcock and Wilcox Company, Lynchburg, Virginia
• OSTI ID: 6291248
• Report Number(s): CONF-840647-
• Conference Information: Journal Name: Am. Soc. Mech. Eng., (Pap.); (United States) Journal Volume: 84-MAT-11
• Country of Publication: United States
• Language: English

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CONTAINERS
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DIAGRAMS
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FAILURE MODE ANALYSIS
FRACTURE MECHANICS
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JOINTS
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TEMPERATURE DEPENDENCE
TENSILE PROPERTIES
US NRC
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WATER MODERATED REACTORS
WELDED JOINTS