Benchmark of Subchannel Code VIPRE-W with PSBT Void and Temperature Test Data

This paper summarizes comparisons of VIPRE-W thermal-hydraulic subchannel code predictions with measurements of fluid temperature and void from pressurized water reactor subchannel and bundle tests. Using an existing turbulent mixing model, the empirical coefficient derived from code predictions in comparison to the fluid temperature measurement is similar to those from previous mixing tests of similar bundle configurations. The predicted steady-state axial void distributions and time-dependent void profiles based on the Lellouche and Zolotar model generally agree well with the test data. The void model tends to predict lower void at the upper elevation under bulk boiling. The void predictions are in closer agreement with the measurements from the power increase, temperature increase, and flow reduction transients than the depressurization transient. Additional model sensitivity studies showed no significant improvement in the code predictions as compared to the published test data. 1. Introduction VIPRE-W (VIPREW or VIPRE) is Westinghouse version of the VIPRE-01 (VIPRE-01 is owned by the Electric Power Research Institute, Palo Alto, CA, USA) thermal-hydraulic subchannel code developed for light water reactor core design applications. The Penn State University (PSU), in cooperation with the Japan Nuclear Energy Safety Organization (JNES), under the sponsorship of the Organization for Economic Co-Operation and Development (OECD) and the United States Nuclear Regulatory Commission (USNRC), has developed benchmark exercises based on the Nuclear Power Engineering Corporation (NUPEC) pressurized water reactor (PWR) subchannel and bundle tests (PSBTs) [1]. The benchmark exercises include Phase I/Exercise 2 for the steady-state void distributions, Phase I/Exercise 3 for the transient void distributions, and Phase II/Exercise 1 for the steady-state fluid temperature distributions in the rod bundles under PWR design conditions. This paper summarizes comparisons of VIPRE-W code predictions with the PSBT fluid temperature and void data, as well as sensitivity studies on the VIPRE-W modeling options. 2. Test Description The PSBT problem specifications [1] provide a description of the NUPEC test facility and rod bundle designs. The test bundles were in configurations for the void and exit temperature measurements, simulating a PWR fuel assembly design with a fuel rod outside diameter (OD) of 9.5？mm containing simple support and mixing vane (MV) grid spacers. The test section for the rod bundle void distribution measurement and an axial diagram indicating locations