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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2016.1
- 수록면
- 1,071 - 1,082 (12page)
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초록· 키워드
오류제보하기
Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.
목차
등록된 정보가 없습니다.
참고문헌 (7)
참고문헌 신청
J.H. Eoh / 2010 / New design options free from a potential sodium freezing issue for a passive DHR system of KALIMER / Nucl. Tech 170:290~305
H.Y. Jeong / 2011 / Thermal-Hydraulic Model in MARS-LMR
American Nuclear Society / 1979 / Decay Heat Power in Light Water Reactors, ANSI/ANS-5.1 / American National Standards Institute
KWI-SEOK HA / 2009 / DEVELOPMENT OF THE MATRA-LMR-FB FOR FLOW BLOCKAGE ANALYSIS IN A LMR / Nuclear Engineering and Technology 41(6):797~806
Toshiba Corporation / 2009 / 4S Safety Analysis, Toshiba Technical Report
H.Y. Jeong / 2011 / Thermal-Hydraulic Model in MARS-LMR
American Nuclear Society / 1979 / Decay Heat Power in Light Water Reactors, ANSI/ANS-5.1 / American National Standards Institute
KWI-SEOK HA / 2009 / DEVELOPMENT OF THE MATRA-LMR-FB FOR FLOW BLOCKAGE ANALYSIS IN A LMR / Nuclear Engineering and Technology 41(6):797~806
Toshiba Corporation / 2009 / 4S Safety Analysis, Toshiba Technical Report
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