• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.629-634
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• 1998

프랑스 CEA 실험장치인 BETHSY 실험설비에서 수행된 부분충수 운전에 대한 실험인 6.9c 실험에 대한 MELCOR 1.8.3 의 평가를 수행하였다. 이 연구는 OECD/NEA 국제공동연구인 ISP-38 로 수행되었다. 평가결과, MELCOR 1.8.3 은 부분충수운전시 잔열제거계통 상실에 대한 예측능력이 있고 원자로 냉각재계통압력, 노심수위 등 전반적으로 거동을 잘 모의하고 있다고 판단되었다. 그러나 민감도 분석에서 도출된 결론에 의하여 상간의 운동량 전달 및 Liquid Entrainment모델에 있어서 개선 필요성이 있는 것으로 평가되었다.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2504-2515
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• 2023

In past years the Paul Scherrer Institute (PSI, Switzerland) and the Karlsruhe Institue of Technology (KIT, Germany)) collaborated to develop a model to account for the active role of nitrogen in the air oxidation of a Zircalloy cladding. The "PSI-KIT Nitriding Model for Zirconium based Fuel Cladding" model was implemented at PSI into PSI-MELCOR 1.8.6. In order to make a preliminary evaluation of the effect of the new model on the evolution of full-scale spent fuel pool accidents, one spent fuel pool event was analyzed using the PSI research version of PSI-MELCOR 1.8.6, which includes the nitriding model. To adapt an existing input deck for the calculations, a sensitivity study was conducted to find an optimal nodalization for the analyses. The nitriding model results were compared to those calculated with the MELCOR 1.8.6-PSI without the new nitriding model. The results demonstrate the effect of the nitriding reactions in spent fuel pool accident progression. Moreover, they confirm the impact of ZrN formation during cladding oxidation in air when the oxidation reactions lead to oxygen starvation inside the fuel assemblies. The nitriding reaction led to higher chemical heat generation during the accident and to an earlier failure of the cladding than when the effect of nitrogen reactions was not considered. It should be noted that the nitriding model, as implemented in the PSI version of MELCOR 1.8.6 has not yet been conclusively validated. Thereby the results presented in this paper should be treated as a preliminary demonstration of the capabilities of the model.

• 에너지공학
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• 제25권2호
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• pp.1-20
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• 2016

원전의 설계기준사고 및 중대사고 해석에서 응축열전달 모델은 매우 중요하며, 특히 피동냉각계통의 개발이 활발히 진행됨에 따라 그 중요성이 더욱 부각되었다. 그런데, 원자로건물 내부에서와 같이 비응축성기체가 존재하는 경우 응축열전달은 현저히 감소하므로 원전 안전해석에서 이를 고려한 응축열전달 모델이 주목받고 있다. 본 연구에서는 냉각재상실사고 등이 발생하는 경우 원자로건물 내부의 상황과 유사한 열수력 조건에서 수행된 응축열전달 실험자료를 이용하여 중대사고 해석코드 MELCOR 1.8.6의 응축열전달 모델을 평가하였다. 실험조건을 응축면의 형상에 따라 네 가지(수직평판, 수직관 외벽, 수직관 내벽, 수평관 내벽)로 분류하였고, 각 분류별 실험들을 MELCOR 코드로 해석하였다. 해석결과, 수직관 내벽을 제외한 나머지 조건에서 MELCOR 코드가 응축열전달을 전체적으로 저 예측하여 개선이 필요한 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3102-3113
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• 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.

• Journal of Radiation Protection and Research
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• 제42권2호
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• pp.106-113
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• 2017

Background: Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. Materials and Methods: The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. Results and Discussion: The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. Conclusion: The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1174-1183
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• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.