Verification of advanced coupled spatial kinetics and thermal hydraulic system codes for accident analysis

Reactor accident scenario simulation provides a means to forecast safety margins and requires a theoretical model of the entire reactor system. Best estimate reactor analysis codes incorporate a full three dimensional (3D) model of a reactor core into a system transient code. The coupled code provides a means to simulate interactions between reactor core behavior and plant dynamics. The Pennsylvania State University (PSU) is verifying the ability of the coupled codes for accident scenario simulations. Verification of the code coupling is a two fold process. The initial verification step tests the feedback between the neutronics and the heat structure components. Next, the feedback between the reactor coolant system and the reactor core is examined. The results of the benchmark are then used to make code to code comparisons. the benchmarks used to verify TRAC-PF1/NEM are the OECD NEACRP rod ejection benchmark and a main steam line break (MSLB) transient at the Three Mile Island Unit 1 (TMI-1) nuclear power plant (NPP). These benchmarks are based on well defined problems with a complete set of input data. The results of the benchmarks are used to validate the reliability of the numerical simulation by following the described verification process