• Proceedings of the Korean Nuclear Society Conference
• /
• 2001.10a
• /
• pp.152.1-152
• /
• 2001

• Nuclear Engineering and Technology
• /
• v.31 no.1
• /
• pp.88-103
• /
• 1999

The purposes of this study are to understand the severe accident phenomena, to establish the simulation method for the experimental test, and to assess the current models in MELCOR for future improvement. This paper presents the results of the PHEBUS FPT1 post test analysis using MELCOR computer code, version 1.8.4. The entire PHEBUS facility has been modeled; the core, the primary circuit including the steam generator, and the containment vessel. Both the thermal hydraulic and the fission product behavior have been investigated. The code simulation results of the thermal hydraulic behavior show good agreement with the experimental data, The fission product release and transport are calculated using the CORSOR models in MELCOR code and the results will be compared with the experiment when the experimental data are available.

• Proceedings of the Korea Society for Simulation Conference
• /
• 2001.05a
• /
• pp.148-152
• /
• 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 Energy Engineering
• /
• v.25 no.2
• /
• pp.1-20
• /
• 2016

A condensation heat transfer model is very important for the safety analysis of nuclear power plants. Especially, condensation under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NCGs in the vapor largely reduces the condensation rate. In this study, the condensation heat transfer model of the severe accident analysis code MELCOR 1.8.6 has been assessed using a set of condensation experiments performed under the thermal-hydraulic conditions similar to those inside a containment during design-basis accidents or severe accidents. Experiment conditions are categorized into 4 types according to the shape of the condensation surface: vertical flat plates, outer surface of vertical pipes, inner surface of vertical pipes, the inner surface of horizontal pipes. The results of the calculations show that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on inner surface of vertical pipes.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.189.1-189
• /
• 2002

• Nuclear Engineering and Technology
• /
• v.32 no.5
• /
• pp.425-432
• /
• 2000

It has been an essential part for the safety assessment of nuclear power plants to understand various phenomena associated with the molten core-concrete interaction(MCCI) under severe accidents. In this study, the severe accident analysis code MELCOR was used to simulate the MCCI experiments such as SWISS and SURC test series which had been performed in Sandia National Laboratories(SNL). The calculation results were compared with corresponding experimental data such as melt temperature, concrete ablation distance, gas generation rate, and aerosol release rate. Good agreements were observed between MELCOR calculation and experimental data. The melt pool was sustained within the range of high temperature and the concrete ablation occurred continuously. The gas generation and aerosol release were under the influence of melt temperature and overlying water pool, respectively.

• Proceedings of the KSME Conference
• /
• 2007.05b
• /
• pp.3491-3496
• /
• 2007

If a molten core is released from a reactor vessel into a reactor cavity during a severe accident, an important safety issue of coolability of the molten core from top-flooding and concrete ablation due to a molten core concrete interaction (MCCI) is still unresolved. The released molten core debris would attack the concrete wall and basemat of the reactor cavity, which will lead to inevitable concrete decompositions and possible radiological releases. In a OECD/MCCI project scheduled for 4 years from 2002. 1 to 2005. 12, a series of tests were performed to secure the data for cooling the molten core spread out at the reactor cavity and for the 2-D long-term core concrete interaction (CCI). The tests included not only separate effect tests such as a melt eruption, water ingression, and crust failure tests with a prototypic material but also 2-D CCI tests with a prototypic material under dry and flooded cavity conditions. The paper deals with the transient simulations on the CCI-2 test by using a severe accident analysis code, CORQUENCH, which was developed at Argonne National Laboratory (ANL). Similar simulations had been already per for me d by using MELCOR 1.8.5 code. Unlike the MELCOR 1.8.5, the CORQUENCH includes a melt eruption mode I and a newly developed water ingression model based on the water ingression tests under the OECD/MCCI project. In order to adjust the geometrical differences between the CCI-2 test (rectangular geometry) and the simulations (cylindrical geometry), the same scaling methodology as used in the MELCOR simulation was applied. For the direct comparison of the simulation results, the same inputs for the MELCOR simulation were used. The simulation results were compared with the previous results by using MELCOR 1.8.5.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.319-320
• /
• 2004

• Nuclear Engineering and Technology
• /
• v.55 no.8
• /
• pp.3102-3113
• /
• 2023

Following the Fukushima nuclear disaster, the simulation of accidents in the spent fuel pool has become more noticeable. Despite the low amount of decay heat power, the consequences of the accidents in a spent fuel pool (SFP) can be severe due to the high content of long-lived radionuclides and lack of protection by the pressure vessel. In this study, the loss-of-cooling accident (LOFA) for the VVER-1000/V446 spent fuel pool is simulated by employing RELAP5 and MELCOR 1.8.6 as the best estimate and severe accident analysis codes, respectively. For two cases with different total power levels, decay heat of spent fuels is calculated by ORIGEN-II code. For modeling SFP of a VVER-1000, a qualified nodalizations are considered in both codes. During LOFA in SFP, the key sequences such as heating up of the pool water, boiling and reducing the water level, uncovering the spent fuels, increasing the temperature of the spent fuels, starting oxidation process (generating Hydrogen and extra power), the onset of fuel melting, and finally releasing radionuclides are studied for both cases. The obtained results show a reasonable consistency between the RELAP5 and MELCOR codes, especially before starting the oxidation process.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05b
• /
• pp.669-674
• /
• 1996

The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction(MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass, The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment.