A Program for Risk Assessment Associated with IGSCC of BWR Vessel Internals

Conference · Sun Aug 01 00:00:00 EDT 1999

OSTI ID:8079

Ware, A G; Morton, D K; Page, J D ^[1]; Nitzel, M E ^[2]; Eide, S A; Chang, T -Y

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1. USNRC
2. INEEL

A program is being carried out for the US Nuclear Regulatory Commission (NRC) by the Idaho National Engineering and Environmental Laboratory (INEEL), to conduct an independent risk assessment of the consequences of failures initiated by intergranular stress corrosion cracking (IGSCC) of the reactor vessel internals of boiling water reactor (BWR) plants. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, both singly and in combination with the failures of others, with specific consideration given to potential cascading and common mode effects on system performance. This paper presents a description of the overall program that is underway to modify an existing probabilistic risk assessment (PRA) of the BWR/4 plant to include IGSCC-initiated failures, subsequently to complete a quantitative PRA.

• Research Organization: Idaho National Lab. (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: US Department of Energy (US)
• DOE Contract Number: AC07-94ID13223
• OSTI ID: 8079
• Report Number(s): INEEL/CON-98-01263; TRN: US0101922
• Resource Relation: Conference: Pressure Vessels and Piping Conference, Boston, MA (US), 08/01/1999--08/04/1999; Other Information: PBD: 1 Aug 1999
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
PERFORMANCE
REACTOR VESSELS
RISK ASSESSMENT
STRESS CORROSION
CRACK PROPAGATION
INTERGRANULAR STRESS CORROSION CRACKING (IGSCC)
REACTOR VESSEL INTERNALS
BOILING WATER REACTOR (BWR) PLANTS
CASCADING AND COMMON MODE EFFECTS
PROBABLISTIC RISK ASSESSMENT (PRA)
GENERAL DESIGN CRITERIA 2 AND 4
NATURAL PHENOMENA
POSTULATED ACCIDENTS
CORE SHROUDS