Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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MNSR transient analysis using the RELAP5/Mod3.2 code

  • Dawahra, S. (Nuclear Engineering Department, Atomic Energy Commission) ;
  • Khattab, K. (Nuclear Engineering Department, Atomic Energy Commission) ;
  • Alhabit, F. (Nuclear Engineering Department, Atomic Energy Commission)
  • Received : 2019.09.17
  • Accepted : 2020.03.10
  • Published : 2020.09.25
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Abstract

To support the safe operation of the Miniature Neutron Source Reactor (MNSR), a thermo-hydraulic transient model using the RELAP5/Mod3.2 code was simulated. The model was verified by comparing the results with the measured and the previously calculated data. The comparisons consisted of comparing the MNSR parameters under normal constant power operation and reactivity insertion transients. Reactivity Insertion Accident (RIA) for three different initial reactivity values of 3.6, 6.0, and 6.53 mk have been simulated. The calculated peaks of the reactor power, fuel, clad and coolant temperatures in hot channel were calculated in this model. The reactor power peaks were: 103 kW at 240 s, 174 kW at 160 s and 195 kW at 140 s, respectively. The fuel temperature reached its maximum value of 116 ℃ at 240 s, 124 ℃ at 160 s and 126 ℃ at 140 s respectively. These calculation results ensured the high inherently safety features of the MNSR under all phases of the RIAs.

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