• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.1075-1076
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.91-92
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.166.1-166
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.67-67
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• 1998

• Journal of computational fluids engineering
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• v.13 no.3
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• pp.21-27
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• 2008

The reactor head of a sodium-cooled fast reactor KALIMER-600 should be cooled during the reactor operation in order to maintain the integrity of sealing material and to prevent a creep fatigue. Analyzing turbulent natural convection flow in the cover gas region of reactor vessel with the commercial CFD code CFX10.0, the cooling requirement for the reactor head and the performance of the insulation plate were assessed. The results showed that the high temperature region around reactor vessel was caused by the convective heat transfer of Helium gas flow ascending the gap between the insulation plate and the reactor vessel inner wall. The insulation plate was shown to sufficiently block the radiative heat transfer from pool surface to reactor head to a satisfactory degree. More than $32.5m^3$/sec of cooling air flow rate was predicted to maintain the required temperature of reactor head.

• Nuclear Engineering and Technology
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• v.33 no.3
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• pp.298-306
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• 2001

As fuel burnup proceeds, thermal gradients, differential swelling, and inter-assembly loading may induce assembly duct bowing. Since duct bowing affects the reactivity, such as long or short term power-reactivity-decrement variations, handling problem, caused by top end deflection of the bowed assembly duct, and the integrity of the assembly duct itself. Assembly duct bowing were first observed at EBR-ll in 1965, and then several designs of assembly ducts and core restraint system were used to accommodate this problem. In this study, NUBOW-2D KMOD was used to analyze the bowing behavior of the assembly duct under the KALIMER(Korea Advanced Liquid MEtal Reactor) core restraint system conditions. The mechanical behavior of assembly ducts related to several design parameters are evaluated. ACLP(Above Core Load Pad) positions, the gap distance between the ducts, and the gap distance between the duct and restraint ring were selected as the sensitivity parameter for the evaluation of duct deflection.

• Nuclear industry
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• v.20 no.6 s.208
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• pp.31-44
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• 2000

본 연구에서 액체금속로의 노심용융(core meltdown)으로 인한 초 즉발 임계(super-prompt critical)의 출력 폭주 사고시, 노심의 반응도 및 열수력 특성 변화와 에너지 방출량등을 계산하기 위하여, Bethe-Tait 방버론을 수정, 보완한 분석 모델이 개발되었다. 주요 보완 내용으로서는, 금속 연료 노심의 단상 액체 영역에서의 선형의(Linear) threshold 형태의 상태 방정식뿐만 아니라 포화 증기(saturated fuel vapor) 영역에서의 상태 방정식이 개발되었고, 이에 따른 노심 붕괴 반응도(disassembly reactivity)의 분석 모델이 개발되었다. 또한 도플러 반응도 효과를 고려하기 위한 분석모델도 아울러 개발되었다. 상기 보완 모델을 실행할 수 있는 수치 해석 프로그램이 개발되었고, 이를 활용하여 KALIMER에서 HCDA가 발생하였을 경우 노심에서의 에너지 방출량 계산이 수행되었다. 분석결과 도플러 효과와 포화 증기 영역에서의 압력 증가 및 노심팽창의 중요성이 확인되었다. 도플러 효과가 고려되지 않을 경우 HCDA는 분석된 모든 반응도 삽입률에 대하여 폭발적인 에너지 방출과 함께 사고가 종결되는 것으로 평가되었다. 그러나 도플러 상수가 최적 평가치인 -0.002인 경우 50$/s이하의 반응도 삽입률에서는 노심은 비등점(0.8KJ/g)에 도달치 않았으며, 설계 기준 사고인 100$/s의 경우에도 노심은 포화 증기 영역에 머물고 압력이 급격히 증가하는 단상(single phase)액체 영역의 threshold 값에 미치지 않기 때문에 사고는 핵연료 증기(vapor)의 점진적인 분산과 함께 종결되는 것으로 분석되며, 총 에너지 발생량은 약 1,800MJ로서 기계적 손상 에너지로 전환되는 분율을 고려할 때 KALIMER 원자로 용기의 구조 설계 기준치에 비해 상당한 여유도를 갖는 것으로 평가되었다.

• Nuclear Engineering and Technology
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• v.33 no.6
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• pp.566-584
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• 2001

In this paper, the progressive deformation and the creep-fatigue damage for the conceptually designed reactor internals of KALIMER(Korea Advanced Liquid MEtal Reactor) are carried out by using the elastic analysis method in the RCC-MR code for normal operating conditions including the thermal load, seismic load (OBE) and dead weight. The maximum operating temperature of this reactor is 53$0^{\circ}C$ and the total service lifetime is 30 years. Thus, the time- dependent creep and stress-rupture effects become quite important in the structural design. The effects of the thermal induced membrane stress on the creep-fatigue damage are investigated with the risk of the elastic follow-up. To calculate the thermal stress, detailed thermal analyses considering conduction, convection and radiation heat transfer mechanisms are carried out with the ANSYS program. Using the results of the elastic analysis, the progressive deformation and creep-fatigue damages are calculated step by step using the RCC-MR in detail. This paper ill be a very useful guide for an actual application of the high temperature structural design of the nuclear power plant accounting for the time-dependent creep and stress-rupture effects.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3181-3188
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• 2008

The supercritical $CO_2$ Brayton cycle energy conversion system is presented as a promising alternative to the present Rankine cycle. The principal advantage of the S-$CO_2$ gas is a good efficiency at a modest temperature and a compact size of its components. The S-$CO_2$ Brayton cycle coupled to a SFR also excludes the possibilities of a SWR (Sodium-Water Reaction) which is a major safety-related event, so that the safety of a SFR can be improved. KAERI is conducting a feasibility study for the supercritical carbon dioxide (S-$CO_2$) Brayton cycle power conversion system coupled to KALIMER(Korea Advanced LIquid MEtal Reactor). The purpose of this research is to develop S-$CO_2$ Brayton cycle energy conversion systems and evaluate their performance when they are coupled to advanced nuclear reactor concepts of the type under investigation in the Generation IV Nuclear Energy Systems. This paper contains the research overview of the S-$CO_2$ Brayton cycle coupled to KALIMER-600 as an alternative energy conversion system.

• Nuclear Engineering and Technology
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• v.32 no.6
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• pp.566-594
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• 2000

In most LMFBR(Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER(Korea Advanced Liquid MEtal Reactor) reactor internal strictures is carried out for normal operating conditions which have the operating temperature 53$0^{\circ}C$ and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME Code Case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects.