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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2017.1
- 수록면
- 979 - 988 (10page)
이용수
초록· 키워드
오류제보하기
The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accidenthas been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during therecirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effectiverecirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop isfeasible if the effective flow area remains above approximately 70% of the nominal flow area. When theflow area is reduced by more than approximately 25% of the nominal value, however, incipience ofboiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, ifa dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperaturein the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would beanticipated. To resolve this, emergency coolant supply through the newly installed external injectionpath can be used as one alternative means of cooling, enabling fuel channel integrity to be maintainedand permanently preventing severe accident conditions. Thus, the external injection flow required toguarantee fuel channel coolability has been estimated.
목차
등록된 정보가 없습니다.
참고문헌 (19)
참고문헌 신청
B. G. Park / 2011 / Experimental study of debris head loss through a pressurized water reactor recirculation sump screen after LOCA / Nucl. Eng. Des 241 : 2462 ~ 2469
E. Krepper / 2010 / CFD-modeling of insulation debris transport phenomena in water flow / Nucl. Eng. Des 240 : 2357 ~ 2364
G. Zigler / 1995 / Parametric study of the potential for BWR ECCS strainer blockage due to LOCA-generated debris : 1-1 ~ 1-7
H. Song / 2014 / Examination of chemical and physical effects on sump screen clogging of containment materials used in Korean plants / Ann. Nucl. Energy 69 : 51 ~ 56
IAEA / 2012 / Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data
E. Krepper / 2010 / CFD-modeling of insulation debris transport phenomena in water flow / Nucl. Eng. Des 240 : 2357 ~ 2364
G. Zigler / 1995 / Parametric study of the potential for BWR ECCS strainer blockage due to LOCA-generated debris : 1-1 ~ 1-7
H. Song / 2014 / Examination of chemical and physical effects on sump screen clogging of containment materials used in Korean plants / Ann. Nucl. Energy 69 : 51 ~ 56
IAEA / 2012 / Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data
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