Thermal Hydraulic Analysis of a Low Enrichment Uranium Core for the MIT Research Reactor

Abstract

The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the MIT reactor is evolving. The objectives of this study are to benchmark the in-house computer code for the MITR, and to perform thermal hydraulic analyses in support of the LEU design studies. The in-house multi-channel thermal-hydraulics code, MULCH-II, was developed specifically for the MITR. This code was validated against PLTEMP for steady-state analysis, and against RELAP5 and temperature measurements for the loss of primary flow transient. The benchmark analysis results showed that the MULCH-II code is in good agreement with other computer codes and experimental data, and hence it is used as the main tool for this study. Various fuel configurations are evaluated as part of the LEU core design optimization study. The criteria adopted for the LEU thermal hydraulics analysis in this study of the limiting safety system settings (LSSS), are to prevent onset of nucleate boiling during steady-state operation, and to avoid a clad temperature excursion during the loss of flow transient. In ranking the LEU core design options, the primary parameter is a low power peaking factor in order to increase the LSSS power and to decrease the maximum clad temperature during the transient. The LEU fuel designs with 15 to 18 plates per element, fuel thickness of 20 mils, and a hot channel factor less than 1.76 are shown to comply with the thermal-hydraulic criteria. The steady-state power can potentially be higher than 6 MW, which was requested in the power upgrade submission to the Nuclear Regulatory Commission.