Abstract
To support the safety analysis of the conversion of the BR2 research reactor to low-enriched uranium (LEU) fuel and extend the validation basis of the RELAP code for the analysis of the conversion of research reactors from highly enriched (HEU) fuel to LEU, the simulation of BR2 tests A/400/1, C/600/3, and F/400/1 was undertaken. These tests are characterized by loss of flow initiated at different reactor power levels with or without loss of system pressure, reactor scram, flow reversal, and reactor cooling by natural circulation. This work presents the RELAP analysis of tests C/600/3 and F/400/1 and comparison of code predictions with experimental measurements for peak cladding temperatures during the transient at different axial locations in an instrumented fuel assembly. The simulations show that accurate representation of the power distribution, especially after reactor scram, between the fuel assemblies and the moderator/reflector regions is critical for the correct prediction of the peak cladding temperatures during the transient. Detailed MCNP and ORIGEN simulations were performed to compute the power distribution between the fuel assemblies and the moderator/reflector regions. With these distributions the predicted peak cladding temperatures are in good agreement with experimental measurements.
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