Safety Analysis of 250-kW Philippine Research Reactor-1 Thermal-hydraulics under Steady-state Operations Using MARS-KS Code

Abstract

Julius Federico M. Jecong1,3,4*, Alvie A. Astronomo2, Frederick C. Hila1, Neil Raymund D. Guillermo1, and Sweng Woong Woo4 1Applied Physics Research Section; 2Nuclear Reactor Operations Section Department of Science and Technology–Philippine Nuclear Research Institute Quezon City, Metro Manila 1101 Philippines 3Nuclear and Quantum Engineering Department Korea Advanced Institute of Science and Technology 291 Daehak-ro, Yuseong-gu, Daejeon 34141 Republic of Korea 4Reactor and Safety Evaluation Department Korea Institute of Nuclear Safety 62 Gwahak-ro, Yuseong-gu, Daejeon 34142, Republic of Korea Safety Analysis of 250-kW Philippine Research Reactor-1 Thermal-hydraulics under Steady-state Operations Using MARS-KS Code Keywords: 250-kW TRIGA reactor, MARS-KS, safety analysis, steady-state, thermal-hydraulics The Philippine Nuclear Research Institute (PNRI) of the Department of Science and Technology (DOST) is implementing a project to use the Philippine Research Reactor-1 (PRR-1) TRIGA nuclear fuel in a subcritical reactor (SCR) for training, education, and research in nuclear science and technology. However, although an SCR is a valuable training and research facility, it offers limited industrial applications. To potentially reap more benefits, DOST-PNRI is also investigating the feasibility of converting the zero-power SCR to a low power critical reactor in the future. In this work, the thermal-hydraulic behavior of a hypothetical 250-kW low power configuration of the PRR-1 TRIGA reactor is investigated using the MARS-KS code. At this power level, the fuel centerline or the hottest region of the fuel was determined to have 236.85- °C temperature during steady-state conditions. This temperature is well below the 749.85-°C temperature limit of TRIGA fuel that will result in excessive swelling. The peak heat flux occurs at the axial center of the fuel rod while the calculated departure from nucleate boiling ratio (DNBR) of 6.18 is minimum at this location. Moreover, the maximum fuel temperature is found to be insensitive to variations in coolant velocity, which ranges from 0.20–0.26 m/s within the natural circulation conditions of this facility. Based on the Bernath critical heat flux (CHF) correlation, coolant velocities below 0.2 m/s will still yield a DNBR above 5.0. These results demonstrate that a 250-kW PRR-1 TRIGA reactor operating under steady-state conditions is capable of safely removing the heat generated in the fuel