RELAP5/MOD3.3 Assessment by Comparison with PKL III G3.1 Experiment (small break in the main steam line) (NUREG/IA-0447)

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Publication Information

Manuscript Completed: April 2014
Date Published: July 2014

Prepared by:
A.Zaruba, K. Krezeminski, A. Benkert

AREVA-NP
Germany.

K. Tien, NRC Project Manager

Division of Systems Analysis
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)

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

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Abstract

The behavior of the PKL facility during a small main-steam-line break was investigated by using the RELAP model of the main facility components, including the reactor pressure vessel, steam generators, pressurizer, primary piping, and main steam lines. The calculations were performed using the computer code RELAP5/MOD3.3.

The main objectives of the study are to compare the simulation results with experimental data and to assess the accuracy of the calculated data. The major parameters and phenomena for simulating a main steam line break, e.g. primary and secondary pressures and temperatures, break flow rate and water level in SG and PZR, were analyzed. The results of the calculations were found to be in good agreement with experimental results.

All the analytical activities were performed as part of the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) projects of Test G3.1.