Anis Bousbia-Salah

Analysis of the Peach Bottom turbine trip 2 experiment by coupled RELAP5-PARCS three-dimensional codes

Abstract Thanks to continuous progress in computer technology, it is now possible to perform best-estimate simulations of complex scenarios in nuclear power plants. This method is carried out through the coupling of three-dimensional (3-D) neutron modeling of a reactor core into system codes. It is particularly appropriate for transients that involve strong interactions between core neutronics and reactor loop thermal hydraulics. For this purpose, the Peach Bottom boiling water reactor turbine trip test was selected to challenge the capability of such coupled codes. The test is characterized by a power excursion induced by rapid core pressurization and a self-limiting course behavior. In order to perform the closest simulation, the coupled thermal-hydraulic system code RELAP5 and 3-D neutron kinetic code PARCS were used. The obtained results are compared to those available from experimental data. Overall, the coupled code calculations globally predict the most significant observed aspects of the transient, such as the pressure wave amplitude across the core and the power course, with an acceptable agreement. However, sensitivity studies revealed that more-accurate code models should be considered in order to better match the void dynamic and the cross-section variations during transient conditions.

Analysis of Rod Ejection Accident in a Research Reactor by the Coupled Technique

This paper investigates the possibility to extend standard computer tools and methods, commonly used in the safety technology of nuclear power reactors, to research reactor safety analysis. A 3-D Neutron Kinetics Thermal-Hydraulic code (3D-NKTH), based on coupling PARCS and RELAP5/3.3 codes, was developed for a standard Material Test Reactor (MTR). The assessment of the model has been performed by comparison of steady state calculations against conventional diffusion codes and Monte Carlo code results. The model is applied for the analysis of a rod ejection accident. The comparison of the 3D-NKTH code, with conventional conservative research reactor tools showed that 3D-NKTH provided a more realistic course of the accident and did not require to define hot channel parameters. This approach could also open new frontiers in the safety analysis of research reactor such as setting realistic safety margin and adequate limits and operation conditions for optimal utilization of research reactors.Copyright