Analysis of the LaSalle Unit 2 Nuclear Power Plant: Risk Methods Integration and Evaluation Program (RMIEP)

Technical Report · Mon Mar 01 00:00:00 EST 1993

OSTI ID:6590372

Lambright, J A; Brosseau, D A; Payne, Jr, A C; Daniel, S L ^[1]

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1. Sandia National Labs., Albuquerque, NM (United States)

This report is a description of the internal fire analysis performed on the LaSalle County Nuclear Generating Station, Unit II. Insights gained from the NUREG-1150 fire analyses of Peach Bottom and Surry were used to the greatest extent possible in this risk assessment. As part of this effort, a new data base for fires was constructed (NUREG/CR-4586). This data base aided in quantification of initiating event frequencies. The most detailed integration with internal events analysis to date in fire risk assessment was also accomplished. The same system fault trees used for internal events were utilized in the fire analysis, which included modeling of components down to the contact pair level. Subsidiary equations were created to map the effects of cable failures and spurious actuations. For all components modeled in the internal event fault trees, associated cables were identified and component and cable locations were traced and mapped into the fault trees. A detailed screening analysis was performed which showed most plant areas had a negligible contribution to fire-induced core damage frequency. A detailed analysis of the fire risk resulted in a total (mean) core damage frequency of 3.21E-5 per year.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution; Sandia National Labs., Albuquerque, NM (United States)
• Sponsoring Organization: USNRC; Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC04-76DP00789
• OSTI ID: 6590372
• Report Number(s): NUREG/CR-4832-Vol.9; SAND-92-0537-Vol.9; ON: TI93011568
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LA SALLE COUNTY-2 REACTOR
RISK ASSESSMENT
AFTER-HEAT REMOVAL
CONTAINMENT SPRAY SYSTEMS
CONTAINMENT SYSTEMS
DEPRESSURIZATION
ECCS
FAILURE MODE ANALYSIS
FAULT TREE ANALYSIS
FIRES
POWER DISTRIBUTION SYSTEMS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
BWR TYPE REACTORS
CONTAINMENT
COOLING SYSTEMS
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
POWER REACTORS
REACTOR PROTECTION SYSTEMS
REACTORS
REMOVAL
SAFETY
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled