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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2020.9
- 수록면
- 1,990 - 1,997 (8page)
- DOI
- 10.1016/j.net.2020.03.006
이용수
초록· 키워드
오류제보하기
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 C at 240 s, 124 C at 160 s and 126 C at 140 s respectively. These calculation results ensured the high inherently safety features of the MNSR under all phases of the RIAs.
목차
등록된 정보가 없습니다.
참고문헌 (14)
참고문헌 신청
2018 / IAEA-TECDOC-1844, Analyses Supporting Conversion of Research Reactors from High Enriched Uranium Fuel to Low Enriched Uranium Fuel, The Case of the Miniature Neutron Source Reactors / VIENNA
E. H. K. Akaho / 2002 / Simulation of reactivity transients in a miniature neutron source reactor core / Nucl. Eng. Des 213:31~42
F. E. Dunn / 2007 / MNSR transient analyses and thermal hydraulic safety margins for HEU and LEU cores using the RELAP5-3D code / International Meeting on Reduced Enrichment for Research and Test Reactors, Czech Republic
C.O. Henry / 2014 / Reactor Core Conversion Studies of Ghana Research Reactor-1 and Proposal for Addition of Safety Rod
S. A. Jonah / 2006 / Criticality calculations and transient analysis of MNSR / Reduced Enrichment for Reduced and Test Reactors
E. H. K. Akaho / 2002 / Simulation of reactivity transients in a miniature neutron source reactor core / Nucl. Eng. Des 213:31~42
F. E. Dunn / 2007 / MNSR transient analyses and thermal hydraulic safety margins for HEU and LEU cores using the RELAP5-3D code / International Meeting on Reduced Enrichment for Research and Test Reactors, Czech Republic
C.O. Henry / 2014 / Reactor Core Conversion Studies of Ghana Research Reactor-1 and Proposal for Addition of Safety Rod
S. A. Jonah / 2006 / Criticality calculations and transient analysis of MNSR / Reduced Enrichment for Reduced and Test Reactors
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