• Journal of Energy Engineering
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• v.25 no.4
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• pp.124-132
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• 2016

In order to estimate the effect of flow distributors connected to an upper nozzle of CMT(Core Makeup Tank) on the thermal-hydraulic characteristics in the tank, a simplified 2 inch Small Break Loss of Coolant Accident(SBLOCA) was simulated by skipping the decay power and Passive Residual Heat Removal System(PRHRS) actuation. The CMT is a part of safety injection systems in the SMART (System Integrated Modular Advanced Reactor). Each test was performed with reliable boundary conditions. It means that the pressure distribution is provided with repeatable and reproducible behavior during SBLOCA simulations. The maximum flow rates were achieved at around 350 seconds after the initial opening of the isolation valve installed in CMT. After a short period of decreased flow rate, it attained a steady injection flow rate after about 1,250 seconds. This unstable injection period of the CMT coolant is due to the condensation of steam injected into the upper part of CMT. The steady injection flow rate was about 8.4% higher with B-type distributor than that with A-type distributor. The gravity injection during hot condition tests were in good agreement with that during cold condition tests except for the early stages.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.604-609
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• 1996

원자력 발전소 부분 충수운전 중에 발생할 수 있는 정지냉각기능 상실사고인 과배수 사건에 대한 확률론적 안전성 평가를 수행하였다. 본 분석의 주된 목적은 과배수로 인한 정지냉각기능 상실사건에 대하여 노심손상 빈도를 계산하고 안전성 향상방안을 도출하는데 있다. 과배수 사건은 초기 부분 충수운전중 발생하는 것으로 가정하였으며 이 때의 발전소 배열(Plant Configuration)은 영광 3,4호기의 운전절차서 및 발전소 운전경험을 근거로 결정하여. 현재 운전상태에 대한 확률론적 안전성평가를 수행하였다. 분석결과 인간오류가 노심손상빈도에 가장크게 기여하는 인자로 나타났으며 인간오류를 줄 일수 있는 대체냉각 절차를 선정하여 재분석을 수행하였다. 고려된 대체냉각 수단은 피동적인 잔열제거 방법인 열규응축냉각(Reflux Cooling)과 정지냉각펌프의 대체계통으로 격납용기 살수펌프를 사용하는 경우의 두가지이다. 본 분석에서는 두가지 대체냉각수단을 모두 채택하는 것으로 가정하여 대체냉각 사용에 따른 효과를 비교하였는데 노심손상 빈도가 1/1000로 감소 하였다. 따라서 절차서 개정에 의한 대체 냉각수단확보는 부분 충수운전중 발전소 안전성 향상에 매우 효과가 큰 것으로 나타났다.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.10
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• pp.1251-1259
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• 2013

In this study, high temperature design and creep-fatigue damage evaluation of a decay heat exchanger (DHX) in the decay heat removal systems of a sodium-cooled fast reactor (SFR) have been performed. Detail design and 3D finite element analysis have been conducted for the DHXs to be installed in active and passive decay heat removal systems in Korean Generation IV SFR, and the DHX installed in the STELLA-1(Sodium integral effect test loop for safety simulation and assessment) at KAERI (Korea Atomic Energy Research Institute). Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two Mod.9Cr-1Mo steel heat exchangers according to the elevated temperature design codes of ASME Section III Subsection NH and RCC-MR code. Code comparisons were made based on the creep-fatigue damage evaluation and issues on conservatisms of the design codes were discussed.