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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2021.9
- 수록면
- 2,775 - 2,787 (13page)
- DOI
- https://doi.org/10.1016/j.net.2021.03.027
이용수
초록· 키워드
오류제보하기
At present, most of the widely used system codes for nuclear safety analysis are one-dimensional, whichcannot effectively simulate the flow field of the reactor core or other structures. This is true even for thesystem codes containing three-dimensional modules with limited three-dimensional simulation function such as RELAP-3D. In contrast, the computational fluid dynamics (CFD) codes excel at providing adetailed three-dimensional flow field of the reactor core or other components; however, the computational domain is relatively small and results in the very high computing resource consuming. Therefore,the development of coupling codes, which can make comprehensive use of the advantages of system andCFD codes, has become a research focus.
In this paper, a review focus on the researches of coupled CFD and thermal-hydraulic system codes wascarried out, which summarized the method of coupling, the data transfer processing between CFD andsystem codes, and the verification and validation (V&V) of coupled codes. Furthermore, a series ofproblems associated with the coupling procedure have been identified, which provide the general direction for the development and V&V efforts of coupled codes.
목차
등록된 정보가 없습니다.
참고문헌 (3)
참고문헌 신청
A. T. Edwards / 1970 / Studies of phenomena connected with the depressurization of water reactors / J. Brit. Nucl. Energy Soc 9(2):125~135
Q. Peng / 2013 / Analytical study on coupling of CATHARE and TRIO_U code for nuclear reactor thermal-hydraulic analysis / Nucl. Power Eng 34(S1):201~205
D. L. Aumiller / 2001 / A coupled RELAP5-3D/CFD methodology with a proof-of-principle calculation / Nucl. Eng. Des 205(1/2):83~90
Q. Peng / 2013 / Analytical study on coupling of CATHARE and TRIO_U code for nuclear reactor thermal-hydraulic analysis / Nucl. Power Eng 34(S1):201~205
D. L. Aumiller / 2001 / A coupled RELAP5-3D/CFD methodology with a proof-of-principle calculation / Nucl. Eng. Des 205(1/2):83~90
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