• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2009년 추계학술대회논문집
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• pp.121-128
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• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.

• 한국산학기술학회논문지
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• 제14권7호
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• pp.3121-3126
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• 2013

원전 배관의 건전성평가는 원자로 안전을 위해 중요하며 결함발견 시 반드시 건전성을 확보해야만 한다. 균열을 갖는 구조물에 대한 정확한 응력확대계수 해석과 균열성장속도는 파괴강도와 피로수명을 평가하는데 필요로 한다. 피로설계와 수명평가는 배관, 산업공장장비 등과 같은 구조물을 설계하는데 극히 중요하다. 응력확대계수를 이용한 균열간의 상호 간섭해석은 유한요소법으로 구하였다. 내압을 받는 원통형구조물의 경우 표면균열의 인접점에서 간섭이 가장 크게 일어남을 확인하였다. 또한, 반복하중 균열에 대해서는 균열 성장평가와 더불어 피로하중에 의한 균열진전을 예측하기 위한 피로해석을 수행하였다.

• 에너지공학
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• 제24권3호
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• pp.135-141
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• 2015

중대사고시 핵연료와 원자로 내부 구조물이 용융되어 원자로용기의 하부에 재배치되면 밀도차이에 의하여 상부의 금속용융물층과 하부의 혼합물층으로 나누어진다. 하부 반구의 혼합물층에서는 지속적으로 붕괴열이 발생하고 이 열은 원자로용기의 건전성을 위협한다. 본 연구는 반구 내부의 체적 열원(Volumetric heat source)이 내재된 매질에서의 자연대류 열전달 현상을 물질전달 실험방법을 이용하여 모사하였다. 황산-황산구리의 구리도금계를 물질전달계로 사용하여 모사를 수행하였다. 수정 Rayleigh 수 $3{\times}10^{14}$에 대하여 Nusselt 수는 반구 하단에서 최소값을 보였고 곡면부를 따라 최상단으로 갈수록 증가하였다.

• Nuclear Engineering and Technology
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• 제34권5호
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• pp.421-432
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• 2002

As one of the advanced design features of the APR1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, test results of direct ECC bypass performed in the steam-water test facility tailed MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) are presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’with the l/4.93 length scale . From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region are obtained.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.559-570
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• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.219-222
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• 2013

Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7-28) ${\times}10^{19}n/cm^2$ (E>0.1MeV) at $250^{\circ}C$, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at $250^{\circ}C$ did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1%) at fluences of (0.7~28) ${\times}10^{19}n/cm^2$ (E>0.1MeV).

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.484-497
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• 2021

This paper proposes an extended time uncertainty analysis approach in Level 2 human reliability analysis (HRA) considering severe accident management (SAM) strategies. The method is a time-based model that classifies two time distribution functions-time required and time available-to calculate human failure probabilities from delayed action when implementing SAM strategies. The time required function can be obtained by the combination of four time factors: 1) time for diagnosis and decision by the technical support center (TSC) for a given strategy, 2) time for strategy implementation mainly by the local emergency response organization (ERO), 3) time to verify the effectiveness of the strategy and 4) time for portable equipment transport and installation. This function can vary depending on the given scenario and includes a summation of lognormal distributions and a choice regarding shifting the distribution. The time available function can be obtained via thermal-hydraulic code simulation (MAAP 5.03). The proposed approach was applied to assess SAM strategies that use portable equipment and safety depressurization system valves in a total loss of component cooling water event that could cause reactor vessel failure. The results from the proposed method are more realistic (i.e., not conservative) than other existing methods in evaluating SAM strategies involving the use of portable equipment.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.283-291
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• 2024

Sodium cooled Fast Reactors (SFR) are built with several engineered safety features and hence a severe accident such as a core melt accident is hypothetical with a probability of <10^-6/ry. However, in case of such accidents, the mixture of the molten fuel and structural materials interacts with sodium. This phenomenon is known as Molten Fuel Coolant Interaction (MFCI) and results in fragmentation of the melt due to various instabilities. The fragmented particles settle as a debris bed on the core catcher at the bottom of the reactor vessel, and continue to generate decay heat. Characteristics of the debris particles play a vital role in heat transfer from the bed and need thorough investigation. The size, shape, and physical state of the debris depend on the associated fragmentation mechanism, superheating of the melt, and sodium temperature. Experiments have been conducted by releasing simulated corium, a molten mixture of alumina and iron generated by the aluminothermy process at ~2400 ℃ into liquid sodium, to study the fragmentation phenomena. After the experiment, the fragmented debris was retrieved and the particle size distribution was determined by sieve analysis. The debris was subjected to microscopic investigation for obtaining morphological characteristics. Based on the characteristics of debris, an attempt has been made to assess of fragmentation mechanism of simulated corium in sodium.

• 비파괴검사학회지
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• 제20권5호
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• pp.390-396
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• 2000

원자로 재료인 SA 508 Class 3 강용접부 및 열영향부 모사 시험편에 대해서 초음파공명분광법으로 동적탄성계수를 측정하였다. 등방성 탄성계수를 가정하여 초기 추정 탄성 계수, $c_{11},\;c_{12}$ 및 $c_{44}$로부터 장방형 시편의 공명 주파수를 계산하였으며 계산된 주파수와 초음파공명분광법으로 측정된 주파수를 비교, 반복 수렴 절차를 거쳐 정밀한 탄성계수를 구했다. 열처리 조건의 차이 및 미세 조직의 차이에 따라 영률 및 전단 계수의 차이가 확실하게 나타났다. 미세한 베이나이트 조직에서의 영률 및 전단 계수는 조대한 마르텐사이트 조직보다 높았으며 이러한 경향은 미세 경도 시험 등의 다른 실험 결과와도 일치하였다.

• Nuclear Engineering and Technology
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• 제23권1호
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• pp.56-65
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• 1991

RELAP5/MOD2 Cycle 36.04코드를 이용하여 LOFT대형냉각재 상실사고 모사실험 L2-3를 계산함으로써 코드의 대형냉각재상실사고에 관련된 열수력현상 예측능력을 평가하였다. 기본계산에서 원자로 압력용기는 이중노심유로와 분리강수관 모델로 모사되었다. 기본계산의 결과 계통의 전반적인 수력학적 거동과 감압기간동안 노심 고출력 부위에서의 열적 거동은 비교적 타당하게 예측되었다. 한편 과냉각-이상유동의 천이 기간동안 임계유량모델, 고질량유속에서의 임계열유속 상관식, 감압기간중의 재접수(Blowdown Rewet)의 판정기준등 코드의 모델/상관식의 부분적 결함이 발견되었다. 이 결함들에 의해 냉각재 재고량이 과대 평가되어 재환수기간의 노심의 열적거동 예측의 정확도가 감소되었다. RELAP5/MOD2 Cycle 36.04로 부터 개선된 코드를 사용한 계산 결과 재접수 현상의 예측 정확도를 개선할 수 있었다.