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학술연구/단체지원/교육 등 연구자 활동을 지속하도록 DBpia가 지원하고 있어요.
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연구자들이 자신의 연구와 전문성을 널리 알리고, 새로운 협력의 기회를 만들 수 있는 네트워킹 공간이에요.
논문 기본 정보
- 자료유형
- 학술대회자료
- 저자정보
- 발행연도
- 2014.4
- 수록면
- 317 - 320 (4page)
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초록· 키워드
오류제보하기
Securing structural integrity of fuel assembly during reactor operation is of utmost importance in order to prevent reactor severe accident like Fukushima nuclear power plant through flow characteristics tests with test assembly scaled down from a prototype reactor of a sodium-cooled fast reactor (SFR) and integrity evaluation under creep-fatigue loads. In this study a preliminary stress analysis for 37-pin test assembly scaled down in terms of simulations on geometry (pith to diameter, lead length to diameter) and flow conditions by Reynolds number from full scale 217-pin sub-assembly in a Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR) of 150 MWe capacity. A full 3-D finite element analysis has been conducted for the 37-pin test assembly assuming that the assembly is to be installed in reactor core to investigate the structural integrity of sub-assembly in PGSFR at the current preliminary design stage. It was shown that the stress intensity range of primary and secondary stress of 3S<SUB>m</SUB> was within design code allowable.
목차
Abstract
1. 서론
2. 노심부수로 시험장치 및 시험집합체
3. 시험집합체의 3D 유한요소 해석
4. 시험집합체의 고온 건전성 평가
5. 결론
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UCI(KEPA) : I410-ECN-0101-2015-500-001458446