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The RELAP/MOD3.2 computer code has been assessed using rod bundle critical heat flux data from the KS-1 and V-200 facilities. This work was performed as part of the U.S. Department of Energy’s International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem 7, these tests addressed one of the important phenomena related to VVER behavior that the code needs to simulate well, core heat transfer. The code was judged to be in minimal agreement with the experiment data, consistently overpredicting the measured critical heat flux. It is recommended that a model development effort be undertaken to develop a critical heat flux model for RELAP5 that better represents the behavior in VVER rod bundles

RELAP5/MOD3.2 Assessment Using CHF Data from the KS-1 and V-200 Experiment Facilities