• Nuclear Engineering and Technology
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• 제31권1호
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• pp.88-103
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• 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.

• Nuclear Engineering and Technology
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• 제32권5호
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• pp.425-432
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• 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.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1143-1156
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• 2017

This paper describes recent development activities of the GENESIS code, which is a transport code for heterogeneous three-dimensional geometry, focusing on applications to reactor core analysis. For the treatment of anisotropic scattering, the concept of the simplified Pn method is introduced in order to reduce storage of flux moments. The accuracy of the present method is verified through a benchmark problem. Next, the iteration stability of the GENESIS code for the highly voided condition, which would appear in a severe accident (e.g., design extension) conditions, is discussed. The efficiencies of the coarse mesh finite difference and generalized coarse mesh rebalance acceleration methods are verified with various stabilization techniques. Use of the effective diffusion coefficient and the artificial grid diffusion coefficients are found to be effective to stabilize the acceleration calculation in highly voided conditions.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.39-46
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• 2014

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1513-1525
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• 2024

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

• Safety and Health at Work
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• 제8권1호
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• pp.42-48
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• 2017

Background: As one of the most frequently occurring accidents in a chemical plant, a fire accident may occur at any place where transfer or handling of combustible materials is routinely performed. Methods: In particular, a jet fire incident in a chemical plant operated under high pressure may bring severe damage. To review this event numerically, Computational Fluid Dynamics methodology was used to simulate a jet fire at a pipe of a compressor under high pressure. Results: For jet fire simulation, the Kemeleon FireEx Code was used, and results of this simulation showed that a structure and installations located within the shelter of a compressor received serious damage. Conclusion: The results confirmed that a jet fire may create a domino effect that could cause an accident aside from the secondary chemical accident.

• Nuclear Engineering and Technology
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• 제47권4호
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• pp.443-453
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• 2015

A hybrid safety injection tank (H-SIT) can enhance the capability of an advanced power reactor plus (APR+) during a station black out (SBO) that is accompanied by a severe accident. It may a useful alternative to an electric motor. The operations strategy of the H-SIT has to be investigated to achieve maximum utilization of its function. In this study, the master logic diagram (i.e., an analysis for identifying the differences between an H-SIT and a safety injection pump) and an accident case classification were used to determine the parameters of the H-SIT operation. The conditions that require the use of an H-SIT were determined using a decision-making process. The proper timing for using an H-SIT was also analyzed by using the Multi-dimensional Analysis of Reactor Safety (MARS) 1.3 code (Korea Atomic Energy Research Institute, Daejeon, South Korea). The operation strategy analysis indicates that a H-SIT can mitigate five types of failure: (1) failure of the safety injection pump, (2) failure of the passive auxiliary feedwater system, (3) failure of the depressurization system, (4) failure of the shutdown cooling pump (SCP), and (5) failure of the recirculation system. The results of the MARS code demonstrate that the time allowed for recovery can be extended when using an H-SIT, compared with the same situation in which an H-SIT is not used. Based on the results, the use of an H-SIT is recommended, especially after the pilot-operated safety relief valve (POSRV) is opened.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2488-2498
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• 2021

The wall film-wise condensation plays an important role in the heat transfer processes of heat exchangers, refrigerators, and air conditioner. In the field of nuclear engineering, steam condensation is often utilized in safety systems to remove the core decay heat under both transient and accident conditions. In particular, passive containment cooling system (PCCS), are designed to ensure containment safety under severe accident conditions. A computational fluid dynamics (CFD) scale analysis has been conducted to calculate the heat transfer rate of the PCCS. However, despite the increase in computing power, there are challenges in the long-term transient simulation of containment using CFD scale codes. In this study, a heat structure coupling between the CFD and system analysis codes was performed to efficiently analyze PCCS. In addition, the component unstructured program for interfacial dynamics (CUPID) was improved to analyze the condensation behavior of ternary gas mixtures. Thereafter, the condensation heat transfer on the primary side was calculated using the improved CUPID and CFD code, whereas that on the secondary side was simulated using MARS. Both the coupled codes were validated against the CONAN facility database. Finally, conjugate heat transfer simulations with wall condensation in the presence of non-condensable gases were appropriately performed.

• Nuclear Engineering and Technology
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• 제16권3호
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• pp.141-154
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• 1984

고리 1호기의 소형파단냉각재 상실사고에 의해 개시된 중대사고 유형과 그 현상에 대할 분석이 제시되었다. 본 해석에서는 KAERI에서 기존 전산코드의 수정.보완된 MARCH 전산코드가 사용되었다. 특히 고리 1호기의 소형파단 LOCA 해석시 수소 거동과 중기과압에 대한 평가 및 그 응답성에 중점을 두고 검토되었으며, 2-loop 발전소 데이타 분석 및 debris-Water 상호작용 모델에 대한 비교 분석이 수행되었다. 제 1부 중대 사고유형 분석결과, 저농도에서 H$_2$ burning이 이루어지는 경우 계속적인 수소 생성으로 인해 반복 수소 spike가 야기 되나, 격납용기 설계압력치 보다낮게 예측되었다. 또한 debris/water 상호작용시 core debris의 입자크기는 첨두압력의 크기에 미치는 영향은 미세하나 첨두압력의 발생시점은 dryout모델사용에 의해서 상당히 지연시키게 되었다. 완전한 노심용융 사고시 수소연소와 증기과압으로부터 예측된 격납용기 최대압력은 격납용기 건전성에 심각한 위협을 초래하지 않는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제37권3호
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• pp.265-272
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• 2005

Severe accidents in nuclear power plants can cause hydrogen-generating chemical reactions, which create the danger of hydrogen combustion and thus threaten containment integrity. For containment analyses, a three-dimensional mechanistic code, GOTHIC-3D has been applied near source compartments to predict whether or not highly reactive gas mixtures can form during an accident with the hydrogen mitigation system working. To assess the code applicability to hydrogen combustion analysis, this paper presents the numerical calculation results of GOTHIC-3D for various hydrogen combustion experiments, including FLAME, LSVCTF, and SNU-2D. In this study, a technical base for the modeling oflarge- and small-scale facilities was introduced through sensitivity studies on cell size and bum modeling parameters. Use of a turbulent bum option of the eddy dissipation concept enabled scale-free applications. Lowering the bum parameter values for the flame thickness and the bum temperature limit resulted in a larger flame velocity. When applied to hydrogen combustion analysis, this study revealed that the GOTHIC-3D code is generally able to predict the combustion phenomena with its default bum modeling parameters for large-scale facilities. However, the code needs further modifications of its bum modeling parameters to be applied to either small-scale facilities or extremely fast transients.