Steady state and transient analyses of MNSR reactor using RELAP5 code

Abstract

Abstract Developing a reliable thermal-hydraulic model of a nuclear reactor is an essential process in the steady state and transient analyses. This paper provides the results of best estimate calculation carried out with reference to Iranian Miniature Neutron Source Reactor (MNSR) using the RELAP5 code. Applying the qualified nodalization and the cross-flow effects are some of the advantages in the present model. Here, various transients including step and ramp reactivity insertions were inspected for safety analysis. The obtained results from the code showed a reasonable agreement with the MNSR Safety Analysis Report (SAR) and existing experimental and reference data.