Semiscale S-NC-02 and S-NC-03 Natural Circulation Tests Performed by RELAP5/MOD3.3 Patch05 (NUREG/IA-0513)

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Publication Information

Manuscript Completed: November 2018
Date Published: May 2019

Prepared by:
Andrej Proše

Jožef Stefan Institute
Jamova cesta 39
SI-1000, Ljubljana, Slovenia

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

Experimental tests S-NC-2 and S-NC-3 were used to assess the RELAP5/MOD3.3 Patch05 computer code. The code developers concluded that it appears the interphase drag model allowed too much liquid to be entrained thus affecting the results of the calculation and that further investigation into the interphase drag model is warranted. The purpose of the present study was to investigate by sensitivity study the influence of interphase drag in the primary system.

The natural circulation experiments were performed in the Semiscale Mod-2A test facility, which is a small-scale model of the primary system of a four-loop Pressurized Water Reactor (PWR). The tests selected were Semiscale natural circulation tests S-NC-02 and S-NC-03. For sensitivity calculations, the latest RELAP5/MOD3.3 Patch05 computer code has been used. The ASCII input deck was obtained in the frame of RELAP5 code distribution for the auto validation purposes. The Symbolic Nuclear Analysis Package (SNAP) graphical user interface animation mask has been created to better understand the influence of varying interphase drag on natural calculated physical phenomena and processes. The results for S-NC-2 test showed that interphase drag coefficient has some influence on the mass flow during natural circulation, but smaller than suspected by code developers. For S-NC-3 test, it was confirmed that equating the height and distribution of the two-phase mixture in calculation within the U-tubes with the experimental condition is essential to predict correctly the degraded heat transfer phenomena.