United States Nuclear Regulatory Commission - Protecting People and the Environment

Analysis of a Loss of Normal Feedwater Transient at the Ringhals-3 NPP Using RELAP5/Mod3.3 (NUREG/IA-0234)

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Publication Information

Manuscript Completed: May 2010
Date Published: July 2010

Prepared by:
J. Bánáti, C. Demaziére, and M. Stålek

Chalmers University of Technology
Department of Nuclear Engineering
S-41296 Gothenburg, SWEDEN

A. Calvo, NRC Project Manager

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

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Availability Notice

Abstract

This report gives an account on the development and validation of the RELAP5/Mod3.3 model of the Ringhals-3 pressurized water reactor against a Loss of Normal Feedwater Transient, which occurred on August 16, 2005. The 3rd unit of Ringhals Nuclear Power Plant comprises a 3-loops Westinghouse design pressurized water reactor on the Swedish West Coast.

At first, the RELAP5 model is presented. All the 157 fuel assemblies are modeled individually in the code input. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops.

The transient was initiated by a malfunction of the feedwater valve at the 2nd steam generator. Consequently, the turbines were tripped and, due to the low level in the SG-2 the reactor was scrammed. Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in safe shutdown conditions. Capabilities of the RELAP5 code were challenged in this transient. The calculated values of the parameters show good agreement with the measured data.

A parametric study was performed In order to evaluate the dependence of the steam generator level on the injected auxiliary feedwater flow. It indicated that the turbine driven auxiliary feedwater pump could possibly inject at a higher flowrate than its nominal value.

The work was performed by the Department of Nuclear Engineering, Chalmers University of Technology in the framework of the Ringhals-3 power uprate project, supported by the Swedish Radiation Safety Authority (SSM). The ultimate goal of this project is to perform independent safety analyses of some limiting transients associated to the power uprate. The work carried out so far was targeted towards the development of state-of-the-art modelling capabilities for the Ringhals-3 unit.

The present validational study is a Swedish contribution to the international Code Assessment and Maintenance Program (CAMP).

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