Pressurized thermal shock probabilistic fracture mechanics sensitivity analysis for Yankee Rowe reactor pressure vessel

Technical Report · Sun Aug 01 04:00:00 EDT 1993

DOI:https://doi.org/10.2172/10180762· OSTI ID:10180762

Dickson, T L; Cheverton, R D; Bryson, J W; Bass, B R; Shum, D K.M.; Keeney, J A ^[1]

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1. Oak Ridge National Lab., TN (United States)

The Nuclear Regulatory Commission (NRC) requested Oak Ridge National Laboratory (ORNL) to perform a pressurized-thermal-shock (PTS) probabilistic fracture mechanics (PFM) sensitivity analysis for the Yankee Rowe reactor pressure vessel, for the fluences corresponding to the end of operating cycle 22, using a specific small-break-loss- of-coolant transient as the loading condition. Regions of the vessel with distinguishing features were to be treated individually -- upper axial weld, lower axial weld, circumferential weld, upper plate spot welds, upper plate regions between the spot welds, lower plate spot welds, and the lower plate regions between the spot welds. The fracture analysis methods used in the analysis of through-clad surface flaws were those contained in the established OCA-P computer code, which was developed during the Integrated Pressurized Thermal Shock (IPTS) Program. The NRC request specified that the OCA-P code be enhanced for this study to also calculate the conditional probabilities of failure for subclad flaws and embedded flaws. The results of this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative influence of a number of key input parameters in the Yankee Rowe PTS analysis and (2) data that can be used for readily determining the probability of vessel failure once a more accurate indication of vessel embrittlement becomes available. This report is designated as HSST report No. 117.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering; Oak Ridge National Lab., TN (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 10180762
• Report Number(s): NUREG/CR--5782; ORNL/TM--11945; ON: TI93040428
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200
36 MATERIALS SCIENCE
360103
DEFECTS
EMBRITTLEMENT
FRACTURE MECHANICS
LOSS OF COOLANT
MATHEMATICAL MODELS
MECHANICAL PROPERTIES
O CODES
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
PRESSURE VESSELS
PRESSURIZATION
PROBABILITY
RADIATION EFFECTS
ROWE YANKEE REACTOR
STEELS
STRESS ANALYSIS
THERMAL SHOCK
WELDED JOINTS