Application of best estimate thermal hydraulic code for safety analysis of high flux research reactor

Abstract

Thermal hydraulic safety analysis of nuclear reactors is performed with an objective to ensure that adequate cooling is provided for the reactor core during normal operation, during anticipated transients and also during design basis accident conditions. Though an established international expertise in relation to computational tools, procedures for their application including Best Estimate (BE) methods and comprehensive experimental database exists for Nuclear Power Plants (NPP), this is lacking for research reactors. The importance of transferring these tools and methods of NPP to research reactors has been initiated by IAEA. However, the ranges of parameters of interest in research reactors are quite different from that of NPP due to large variation of their fuel composition, system pressure, materials and overall system configuration. So far, conservative computational tools are used to perform safety analyses for design of these reactors. The application of BE methods has become a real necessity in the safety analysis and design of research reactors. This allows getting more realistic simulation of the processes taking place during the steady state operation and transients. In comparison to the conservative approaches, the application of Best Estimate methods result in a precise prediction of the system behaviour leading to mitigation of constraining limits in design and operation. The purpose of the present paper is to provide an overview of the thermal hydraulic safety analyses carried out for the design of high flux research reactor using best estimate thermal hydraulic code RELAP5. Various parameters such as fuel and clad temperature distribution, ONB temperature and DNBR under all anticipated/operational transients and design basis accidents are presented.