Comparative study of STCP and SCDAP simulation of PBF SFD test 1-1

Over the last decade, there has been substantial development and progress in understanding the nature of severe accidents in light water reactors. Of particular interest is being able to predict the fission product release and transport after the reactor core is damaged. As part of this continuing effort, the US Nuclear Regulatory Commission sponsored the development of the Source Term Code Package (STCP), which models core degradation and fission product release to the environment. In order to better establish the validity of source term predictions from these phenomenological models, extensive benchmarking against experimental data and more detailed mechanistic models are necessary. The purpose of this paper is to present the results of a detailed comparison of the STCP and the SCDAP computer codes for simulation of the Power Burst Facility (PBF) Severe Fuel Damage (SFD) test 1-1. The SCDAP code is mechanistic and has been benchmarked against a broad range of severe accident data. The SFD 1-1 test was designed to simulate the heatup and resulting fuel damage in the upper half of a 3000-MW(thermal) pressurized water reactor core ∼ 2 to 3 h after initiation of a small-break accident when the core is ∼ 75% uncovered