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하귀석,김원석,장원표,류건중,Ha, G.S.,Kim, W.S.,Chang, W.P.,Yoo, K.J. Korean Nuclear Society 1995 Nuclear Engineering and Technology Vol.27 No.6
본 연구는 가압경수로의 부분충수 운전중 잔열제거기능 상실사고 해석시 신뢰성을 확보하기 위해 RELAP5/MOD3.1 코드로 관련 대형 실험을 모의 계산하여, 사고시 예상되는 주요 물리적 현상의 파악과 코드의 예측능력을 평가하는 것이다. 대상 실험으로 선택된 BETHSY Test 6.9a는 이 사고중 증기발생기가 작동하지 않고, 가압기 Manway를 개방한 상태 (Configuration)를 모의한 실험이다. 이 연구 결과는 실제 원전 사고시 예상되는 중요 현상 뿐 아니라, 이에 영향을 미치는 민감한 인자를 파악하여 사고 해석결과의 유효성을 판단하는 데 상당히 기여할 것으로 기대한다. 연구결과 RELAP5/MOD3.1 코드는 대체적으로 계통의 과도기 거동은 타당하게 예측하고 있지만, 모의계산에서 Time-Step이 아주 짧아 막대한 시간이 소요된다는 문제점이 발견되었다. 그 외에도 노심팽창수위 (swollen level)를 과대평가하여 가압기의 수위 및 계통의 압력을 높게 계산하였다. 이로 인해 가압기를 통한 방출량도 과대계산하여 노심노출을 약 400초 빨리 예측하였다. 실험과 코드 예측결과를 종합할 때 가압기 Manway 만의 개방으로는 계통압력이 상승하고, 중력주입냉각수로는 노심수위 회복에 불충분하며, 결국 강제주입에 의해서 노심수위가 회복될 수 있음이 입증되었다. The present study is to understand the physical phenomena anticipated during the accident with RHR loss under mid-loop operation in a PWR and, at the same time, to examine the prediction capability of RELAP5/MOD3.1 on such an accident, by simulating an integral test relevant to this accident for reliable analysis in an actual PWR. The selected experiment, i.g. BETHSY Test 6.9a, represents the configuration with the pressurizer manway open and steam generators unavailable during the accident. Accordingly, the results of this ok are sure to contribute to understanding both the key events as well as the sensitive parameters, anticipated in the accident, for validity of the actual analysis. In the simulation result overall behavior as well as major phenomena observed in the experiment have been predicted reasonably by RELAP5/MOD3.1, however, the problem associated with enormous computing time .due to small time step size has been encountered. Besides, the code prediction of higher swollen level in the pressure vessel has given rise to overestimation of both pressurizer level and RCS pressure. Subsequently, overprediction of the break flow through the manway has led to earlier core uncovery than that in the experiment by about 400 seconds. As a whole, it is demonstrated from both the experiment and the analysis that gravity feed has not been sufficient to recover the core level and thus additional forced feed has been necessary in this configuration.
A Preliminary Safety Analysis for the Prototype Gen IV Sodium-Cooled Fast Reactor
이귀림,하귀석,정재호,최치웅,Taekyeong Jeong,Sang June Ahn,이승원,WON-PYO CHANG,강석훈,유재운 한국원자력학회 2016 Nuclear Engineering and Technology Vol.48 No.5
Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.
이용범,정해용,조충호,권영민,하귀석,장원표,석수동,한도희 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.8
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.
THREE-DIMENSIONAL FLOW PHENOMENA IN A WIRE-WRAPPED 37-PIN FUEL BUNDLE FOR SFR
정재호,유진,이귀림,하귀석 한국원자력학회 2015 Nuclear Engineering and Technology Vol.47 No.5
Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel assembly mock-up of aJapanese loop-type sodium-cooled fast reactor, Monju, were investigated with a numericalanalysis using a general-purpose commercial computational fluid dynamics code, CFX. Complicated and vortical flow phenomena in the wire-wrapped 37-pin fuel assembly werecaptured by a Reynolds-averaged NaviereStokes flow simulation using a shear stresstransport turbulence model. The main purpose of the current study is to understand thethree-dimensional complex flow phenomena in a wire-wrapped fuel assembly to supportthe license issue for the core design. Computational fluid dynamics results show goodagreement with friction factor correlation models. The secondary flow in the corner andedge subchannels is much stronger than that in an interior subchannel. The axial velocityaveraged in the corner and edge subchannels is higher than that averaged in the interiorsubchannels. Three-dimensional multiscale vortex structures start to be formed by aninteraction between secondary flows around each wire-wrapped pin. Behavior of the largescalevortex structures in the corner and edge subchannels is closely related to the relativeposition between the hexagonal duct wall and the helically wrapped wire spacer. Thesmall-scale vortex is axially developed in the interior subchannels. Furthermore, a drivingforce on each wire spacer surface is closely related to the relative position between thehexagonal duct wall and the wire spacer.
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.
ADVANCED SFR DESIGN CONCEPTS AND R&D ACTIVITIES
한도희,장진욱,김영인,김영일,이찬복,김성오,이재한,하귀석,김병호,이용범 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.4
In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical CO2 Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.