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Young-Min KWON(권영민),Kwi-Seok HA(하귀석),Hae-Yong JEONG(정해용),Won-Pyo CHANG(장원표) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.11
The numerical simulation of a 271-rod fuel assembly of nuclear Sodium-cooled Fast Reactor (SFR) with an internal blockage has been carried out. Internal blockage within a subassembly is addressed in the safety assessment because it has potentially very serious consequences for the reactor as a whole. Three dimensional calculations were performed using the SABRE4 computer code for the range of blockage positions and sizes to investigate the seriousness and detectability of the internal blockage. The magnitude and location of the peak temperatures together with the temperature distribution at the subassembly exit were calculated in order to look at the potential for damage within the subassembly, and the possibility of blockage detection. The analysis result shows that the 6-subchannel blockage causes large temperature rise within an assembly with practically no change in the mixed mean temperature at the assembly exit.
김민기(Min Gi Kim),민병채(Byung Chae Min),하귀석(Kwi Seok Ha),배성원(Sung Won Bae),최경민(Gyung Min Choi),정재준(Jae Jun Jeong) 대한기계학회 2020 大韓機械學會論文集B Vol.44 No.6
원자력 열수력 계통 코드 MARS는 원자력발전소의 설계 및 안전성 분석을 위해 개발된 것이다. 이 코드는 다양한 기기로 구성된 열수력 계통의 2상 유동을 일반적으로 모델링할 수 있는 장점이 있으며, 원자력 계통의 정상 및 과도 상태 2상 유동(two-phase flow) 거동 예측에 탁월한 성능을 갖고 있다. 그런데, MARS 코드는 원자력발전소 계통 해석을 위해 개발되었으므로 냉동 사이클에 바로 적용할 수는 없다. 본 연구에서는 MARS 코드를 냉동 사이클 해석에 활용하기 위해 먼저 냉매(R-410A)의 물성치와 압축기 모델을 MARS 코드에 구현하였다. 그 다음, 수정된 MARS 코드로 R-410A의 비등 및 응축 열전달 실험을 해석하여 코드의 열전달 예측 성능을 평가하였다. 이어서 MARS 코드로 멀티 시스템 에어컨 실험을 모의하여 냉동 사이클 적용성을 고찰하였다. 평가 결과, MARS 코드의 냉동 사이클 적용성을 보였으며 아울러 향후 개선 사항 등을 도출하였다. The MARS code is a thermal-hydraulic system analysis code developed for the design and safety analysis of a water-cooled nuclear power plant. It can realistically predict two-phase flow behaviors in a complicated thermal-hydraulic system and can be used for both steady-state and transient calculations. In this work, to utilize the advantages of the MARS code for the refrigeration cycle analysis, we have implemented the thermodynamic property of a refrigerant (R-410A) and a compressor model into the MARS code. Subsequently, we have assessed the MARS model, which was originally developed for water, for boiling and condensation of the refrigerant. Furthermore, we have evaluated the code using experimental data from a multi-system air-conditioner. It is shown that MARS can predict transient behaviors of the refrigeration cycle reasonably well. The limitations of the code and improvements to be made have also been identified.
Impact of Multi-dimensional Core Thermal-hydraulics on Inherent Safety of Sodium-Cooled Fast Reactor
Young-Min KWON(권영민),Hae-Yong JEONG(정해용),Kwi-Seok HA(하귀석) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.11
A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.