United States Nuclear Regulatory Commission - Protecting People and the Environment

Sensitivity Analyses of a Hypothetical 6 Inch Break, LOCA in Ascó NPP using RELAP/MOD3.2 (NUREG/IA-0240)

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Publication Information

Manuscript Completed: August 2010
Date Published: September 2010

Prepared by:
R. Pericas, L. Batet, and F. Reventós

1Institute of Nuclear Energy Research
Atomic Energy Council, R.O.C.
1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325
TAIWAN

Department of Physics and Nuclear Engineering
Technical University of Catalonia
ETSEIB, Av. Diagonal 647, Pav. C
08028 Barcelona
Spain

Antony Calvo, NRC Project Manager

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
Under the International Code Assessment and Maintenance Program (CAMP)

Prepared for:
Division of Systems Analysis
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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Abstract

A postulated 6 inch break Loss-of-Coolant Accident (LOCA) is analyzed for the Ascó I plant (a 3 loop PWR Westinghouse design) using RELAP5/Mod3.2. Different scenarios are calculated, including the total loss of the High Pressure Injection System (HPIS). In this case, the maximum cladding temperature rises above the steady state value at full power but, before the end of the transient, temperatures return to normal values by the effect of the accumulators injection.

Passive heat structures in the reactor pressure vessel have been incorporated to the model (i.e. the pressure vessel walls and some internals). Calculations with the model including passive heat structures lead to more severe scenarios (i.e. higher cladding temperatures) but still accumulators action is able to return the temperatures to normal values.

Finally, sensitivity to the trip time of the reactor coolant pumps is analyzed. Results show that lower cladding temperatures result when the pump trip is delayed.

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