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.