DEVELOPMENT OF A TWO-DIMENSIONAL THERMOHYDRAULIC HOT POOL MODEL AND ITS EFFECTS ON REACTIVITY FEEDBACK DURING A UTOP  IN LIQUID METAL REACTORS

이용범; 정해용; 조충호; 권영민; 하귀석; 장원표; 석수동; 한도희

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자료유형: 학술저널

저자정보: 이용범 정해용 조충호 권영민 하귀석 장원표 석수동 한도희

저널정보: 한국원자력학회 Nuclear Engineering and Technology Nuclear Engineering and Technology 제41권 제8호

발행연도: 2009.1

수록면: 1,053 - 1,064 (12page)

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연구방법

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연구결과

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The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

#Hot Pool Model #Reactivity Feedback #Control Rod Drive Line #Unprotected Transient over Power

참고문헌 (13)

참고문헌 신청

Agrawal, A. K. / 1973 / An Advanced Thermohydraulic Simulation Code for Transients in LMFBRs (SSC-L Code)

CFX Ltd / 2003 / CFX Version 5.6 User’s Manual

Cahalan, J. E. / 1994 / Advanced LMR Safety Analysis Capabilities in the SASSYS- 1 and SAS4A Computer Codes / Proceedings of the Int. Topical Meeting on Advanced Reactor Safety

Chang, W.P. / 2002 / Model Development for Analysis of the Korea Advanced Liquid Metal Reactor / Nuclear Engineering and Design 217 : 63 ~ 80

DOHEE HAHN / 2007 / CONCEPTUAL DESIGN OF THE SODIUM-COOLED FAST REACTOR KALIMER-600 / Nuclear Engineering and Technology 39 (3) : 193 ~ 206

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