RELAP-5/MOD 3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility
The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. 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 Sovietdesigned reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to
... enomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior. iv v CONTENTS ABSTRACT.6 CODE AND INPUT MODEL DESCRIPTION The RELAP5/MOD3.2 code and input model used for the assessment calculations are described below. Also addressed are the initial and boundary conditions used for the transient calculations. RELAP5/MOD3.2 The RELAP5/MOD3.2 computer code 3 was developed at the Idaho National Engineering and Environmental Laboratory (INEEL) for use in analyzing transients in light water reactors. It can be used for simulating a wide variety of system transients of interest in reactor safety. The core, primary system, secondary system, feedwater train, and system controls can be simulated. RELAP5/MOD3.2 uses a one-dimensional, two fluid, nonequilibrium, six equation hydrodynamic model with a simplified capability to treat multi-dimensional flows. This model provides continuity, momentum, and energy equations for both the liquid and the vapor phases within a control volume. The energy equation contains source terms which couple the hydrodynamic model to the heat structure conduction model by a convective heat transfer formulation. The code contains special process models for critical flow, abrupt area changes, branching, crossflow junctions, pumps, accumulators, valves, core neutronics, and control systems. A countercurrent flow limitation model can also be applied at vertical junctions. The reflood model does not work properly in this version of the code.