Simulation of LOCA 6" and LOCA 2" Transients in the RHR of a PWR Under Low Power Conditions Using RELAP5/MOD3.2 (NUREG/IA-0171)

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Publication Information

Date Published: February 2000

Prepared by:
S. Martorell, S. Carlos, G. Verdú, V. Serradell, J. Rodenas, R. Miró, C. Llopis*

*C.N. Vandellos II
Travessera de les Corts, 55, Lateral
08028 Barcelona

Department of Chemical and Nuclear Engineering
Polytechnic University of Valencia
Camino de Vera, 14

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

Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Availability Notice


The present study consists of the simulation of two loss of coolant accidents, LOCA 6" and LOCA 2", in one of the residual heat removal system (RHR) lines outside the containment, using the thermal-hydraulic code RELAP5/MOD3.2.

Both transients have been simulated on a typical three loop, Westinghouse design, pressurized water reactor plant working under shutdown conditions.

The study was focused on the simulation of the most important thermal-hydraulic parameters in order to check the validity of the success criteria assumed in the plant probabilistic safety analysis (PSA) under shutdown conditions. Also to analyze the code capability for simulating shutdown conditions was of interest in this study.

As a result of this study, it can be concluded that the main thermal-hydraulic plant features follow what is foreseen in the plant PSA, although it can not be assured that the values reached are the correct ones due to the lack of experimental data.

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