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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2020.9
- 수록면
- 1,963 - 1,973 (11page)
- DOI
- 10.1016/j.net.2020.03.005
이용수
초록· 키워드
오류제보하기
Liquid metal-cooled reactors use various passive safety systems driven by natural circulation. Investigating these safety systems experimentally is more advantageous by using a simulant. Although numerous experimental approaches have been applied to natural circulation-driven passive safety systems using simulants, there has been no clear validation of the similarity law. To validate the similarity law experimentally, SINCRO-V experiment was conducted using Wood's metal and water for simulant of the Wood's metal. A pair of SINCRO-V facilities with length-scale ratio of 14.1:1 for identical Bo’ was investigated, which was the main similarity parameter in temperature field simulation. In the experimental range of 0.2e1.0% of decay heat, the temperature distribution characteristics of the small water facility were very similar to that of the large Wood's metal facility. The temperature of the Wood's metal predicted by the water experiment showed good agreement with the actual Wood's metal temperature.
Despite some error factors like discordance of Gr’ and property change along the temperature, the water experiment predicted the Wood's metal temperature with an error of 27%. The validity of the similarity law was confirmed by the SINCRO-V experiments
목차
등록된 정보가 없습니다.
참고문헌 (32)
참고문헌 신청
E. E. Feldman / 1987 / EBR-II unprotected loss-of-heat-sink predictions and preliminary test results / Nucl. Eng. Des 101:57~66
H. P. Planchon / 1987 / Implications of EBR-II inherent safety demonstration test / Nucl. Eng. Des 101:75~90
W. K. Lehto / 1987 / Safety analysis for the loss-of-flow and loss-of-heat sink without scram tests in EBR-II / Nucl. Eng. Des 101:35~44
F. Yamada / 2014 / Development of natural circulation analyticalmodel in super-COPD code and evaluation of core cooling capability in monju during a station blackout / Nucl. Technol 188:292~321
T. Ishizu / 2005 / An evaluation of passive safety features of the Japanese prototype LMFBR monju / Proceedings of the 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11)
H. P. Planchon / 1987 / Implications of EBR-II inherent safety demonstration test / Nucl. Eng. Des 101:75~90
W. K. Lehto / 1987 / Safety analysis for the loss-of-flow and loss-of-heat sink without scram tests in EBR-II / Nucl. Eng. Des 101:35~44
F. Yamada / 2014 / Development of natural circulation analyticalmodel in super-COPD code and evaluation of core cooling capability in monju during a station blackout / Nucl. Technol 188:292~321
T. Ishizu / 2005 / An evaluation of passive safety features of the Japanese prototype LMFBR monju / Proceedings of the 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11)
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