• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.363-368
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• 1996

경수로형 원자로의 중대사고 진행시 압력용기내 노심의 용융현상 및 재배치과정 등에 대한 MELCOR 코드내 노심손상모델의 예측 및 분석능력을 검증하고자 하였다. 이를 위하여 노심손상 모의실험중 하나인 독일의 KfK에서 실시된 CORA-13 실험을 선정한 후 이 실험을 MELCOR 1.8.2 코드를 이용하여 계산하였다. 실험결과와 계산결과를 비교분석하고 또한, MELCOR 코드에 대한 민감도분석을 수행함으로써 MELCOR 코드내 손상된 노심의 거동에 대한 열수력모델들을 검증하였다.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1508-1515
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• 2022

In the present study, the cavity module of the MELCOR code is used for the simulation of molten corium concrete interaction (MCCI) during the late phase of postulated large break loss of coolant (LB-LOCA) accident in the APR1400 reactor design. Using the molten corium composition data from previous MELCOR Simulation of APR1400 under LB-LOCA accident, the ex-vessel phases of the accident sequences with long-term MCCI are recalculated with stand-alone cavity package of the MELCOR code to investigate the impact of water ingression and melt eruption models which were hitherto absent in MELCOR code. Significant changes in the MCCI behaviors in terms of the heat transfer rates, amount of gases released, and maximum cavity ablation depths are observed and reported in this study. Most especially, the incorporation of these models in the new release of MELCOR code has led to the reduction of the maximum ablation depth in radial and axial directions by ~38% and ~32%, respectively. These impacts are substantial enough to change the conclusions earlier reached by researchers who had used the older versions of the MELCOR code for their studies. and it could also impact the estimated cost of the severe accident mitigation system in the APR1400 reactor.

• Proceedings of the Korea Society for Simulation Conference
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• 2001.05a
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• pp.148-152
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• 2001

본 논문에서는 중대사고 해석코드인 MELCOR를 이용하여 개발중인 중대사고 훈련용 그래픽 시뮬레이터 MEL-GRS에 대하여 기술하였다. MEL-GRS는 SL-GMS 그래픽 튤과 MELCOR 계산 결과를 적절히 사용하여 중대사고 현상을 실시간으로 디스플레이하는 목적으로 개발되었는데, 기존의 MELCOR 코드에서 불가능했던 다이내믹 시뮬레이션 기능을 가지고 있어 실시간 밸브 및 펌프 조작이 가능하다. 개발된 시스템은 IBM PS Windows 환경에서 작동하며, 울진 3, 4호기를 대상으로 한 TLOFW, LOCA등의 중대사고 시나리오를 사용하여 그 성능을 검증하였다. 개발된 시스템은 차후 발전소 현장의 설치 및 검증을 거쳐 운전원 및 TSC 요원의 중대사고 훈련도구로 활용한 계획이다.

• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.192-204
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• 2012

For a severe accident of nuclear power plant, an approach to estimation of the radiological source term using a severe accident code(MELCOR) has been proposed. Although the MELCOR code has a capability to estimate the radiological source term, it has been hardly utilized for the radiological consequence analysis mainly due to a lack of understanding on the relevant function employed in MELCOR and severe accident phenomena. In order to estimate the severe accident source term to be linked with the radiological consequence analysis, this study proposes 4-step procedure: (1) selection of plant condition leading to a severe accident(i.e., accident sequence), (2) analysis of the relevant severe accident code, (3) investigation of the code analysis results and post-processing, and (4) generation of radiological source term information for the consequence analysis. The feasibility study of the present approach to an early containment failure sequence caused by a fast station blackout(SBO) of a reference plant (OPR-1000), showed that while the MELCOR code has an integrated capability for severe accident and source term analysis, it has a large degree of uncertainty in quantifying the radiological source term. Key insights obtained from the present study were: (1) key parameters employed in a typical code for the consequence analysis(i.e., MACCS) could be generated by MELCOR code; (2) the MELOCR code simulation for an assessment of the selected accident sequence has a large degree of uncertainty in determining the accident scenario and severe accident phenomena; and (3) the generation of source term information for the consequence analysis relies on an expert opinion in both areas of severe accident analysis and consequence analysis. Nevertheless, the MELCOR code had a great advantage in estimating the radiological source term such as reflection of the current state of art in the area of severe accident and radiological source term.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.637-642
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• 1996

This paper presents the analysis results for the core degradation processes and the fission product release of the PHEBUS FPT0 experiment using MELCOR1.8.3. The objective of this study is to assess models associated with the core damage and fission product behavior in MELCOR. The calculation results were much improved through sensitivity studies. Thermal/hydraulic behavior in the core and the circuit was well predicted under the intact core geometry. In non-eutectic model case. the UO$_2$ dissolution model in the MELCOR always showed such a tendency that the resulting dissolved UO$_2$ mass was small at the highly oxidized condition due to the model logic. Total H$_2$ generation mass was underpredicted because the stiffner was not modeled and the liner in the shroud was not allowed to be oxidized in MELCOR. Some difficulties were found in modeling the activation product were solved by manipulating the RN input associated with the initial fission product inventory. These problem were occurred because there are no control rod model in MELCOR. Generally the fission product release ratio showed a similar trend compared with the measured data except the activation product. which have no model to simulate in MELCOR.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.616-621
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• 1996

PWR 원전의 중대사고시 격납건물내 수소거동을 모의한 NUPEC의 수소분포실험 결과를 MELCOR 1.8.3 코드를 이용하여 검증 차원의 비교분석을 수행하였다. 이 연구에서는 정확한 실험조건 및 코드의 특성을 반영하여 실험에서의 유동 및 열역학적 조건을 모두 모의하였다. 이를 통해 실험에서 나타난 수소거동 특성을 재확인하고, MELCOR 코드의 분석능력 및 특성을 평가하였다. ISP-35에 대한 비교분석을 통해 다격실 격납건물내 중대 사고시 수소 혼합 및 분포 현상에 대한 MELCOR의 분석능력을 확인하였다.

• Nuclear Engineering and Technology
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• v.35 no.4
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• pp.286-298
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• 2003

A steam explosion module, STX, has been developed using the mechanistic steam explosion analysis code, TEXAS-V, in order to estimate the dynamic load with steam explosion by implementing the module to the integrated safety analysis code, MELCOR. One of the difficulties in using mechanistic steam explosion codes is that they do not have any obvious criteria for defining some uncertain parameters such as triggering timing, triggering magnitude, mesh axial length and mesh cross-sectional area. These parameters have been user decision parts in the past. Steam explosion sample calculations and sensitivity studies on uncertain parameters were conducted to investigate those uncertain parameters. The TEXAS-V simulations were summarized in the format of a look-up table and a linear interpolation technique was adopted to calculate the steam explosion load between the data points in the table. The STX-module merged with MELCOR showed the same results as the original MELCOR and additionally it could estimate the steam explosion load in the reactor cavity.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1939-1950
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• 2019

The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.225.2-225
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.215-215
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• 1999