Analysis of RELAP5/MOD3.3 Prediction of 2-Inch Loss-of-Coolant Accident at Krško Nuclear Power Plant (NUREG/IA-0222)

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Publication Information

Manuscript Completed: December 2005
Date Published: April 2010

Prepared by:
Iztok Parzer, Borut Mavko
Jožef Stefan Institute
Jamova cesta 39
SI-1000 Ljubljana, Slovenia

A. Calvo, NRC Project Manager

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

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Abstract

The purpose of this analysis was to perform calculations of the loss-of-coolant accident (LOCA) for simulator verification and validation and to study the thermal-hydraulic response of the reactor coolant system.

For the thermal-hydraulic analysis, the RELAP5/MOD3.3 code and input model provided by Krško Nuclear Power Plant was used. A small-break LOCA scenario was analyzed to estimate plant response to the opening of a break in cold leg No. 2 between the reactor coolant pump and the reactor pressure vessel. For the purpose of the analysis, the equivalent diameter of the cross-sectional area of the break was set to 5.08 centimeters (2 inches).

In the presented study, the 2-inch LOCA scenario for the Krško Nuclear Power Plant was analyzed with regard to the differences between the Henry-Fauske and the Ransom-Trapp critical flow model. In addition, the study investigated the effect of the special offtake flow model at the break. Some variation cases were also run to capture the effect of flow bypasses in the reactor vessel on the loop seal clearing phenomena.