• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1210-1216
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• 2018

Site or multi-unit (MU) risk assessment has been a major issue in the field of nuclear safety study since the Fukushima accident in 2011. There have been few methods or experiences for MU risk assessment because the Fukushima accident was the first real MU accident and before the accident, there was little expectation of the possibility that an MU accident will occur. In addition to the lack of experience of MU risk assessment, since an MU nuclear power plant site is usually very complex to analyze as a whole, it was considered that a systematic method such as probabilistic safety assessment (PSA) is difficult to apply to MU risk assessment. This paper proposes a new MU risk assessment methodology by using the conventional PSA methodology which is widely used in nuclear power plant risk assessment. The logical failure structure of a site with multiple units is suggested from the definition of site risk, and a decomposition method is applied to identify specific MU failure scenarios.

• 한국기술혁신학회:학술대회논문집
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• 한국기술혁신학회 2001년도 춘계학술대회:발표자료
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• pp.389-408
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• 2001

Evaluating the results of National Nuclear R&D projects has critical importance in nuclear management aspect. This paper uses multi-attribute utility theory as a basis for obtaining a value Index to assess the results of nuclear R&D projects and applies the theory to a specific Korean case study. To structure and quantify basic values for the . evaluation, we elicited important attributes, then refined and structured them into a hierarchy. A multi-attribute index is constructed as a multi-attribute utility function, based on value judgments provided by a group of technical experts, policy makers, and decision-makers. The implications of the results are also discussed. We found that the work and results can provide valuable insights for assessment of nuclear R&D consequences.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1284-1288
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• 2021

In this work, we propose and design a GaN-based diode with a p-doped GaN (p-GaN) multi-well structure for high efficiency betavoltaic (BV) cells. The short-circuit current density (J_SC) and opencircuit voltage (V_OC) of the devices were investigated with variations of parameters such as the doping concentration, height, width of the p-GaN well region, well-to-well gap, and number of well regions. The J_SC of the device was significantly improved by a wider depletion area, which was obtained by applying the multi-well structure. The optimized device achieved a higher output power density by 8.6% than that of the conventional diode due to the enhancement of J_SC. The proposed device structure showed a high potential for a high efficiency BV cell candidate.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2268-2276
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• 2021

Spent nuclear fuel, which can degrade during long-term storage, must be transported intact in normal transport conditions. In this regard, many studies, including those involving Multi-Modal Transportation Test (MMTT) campaigns, have been conducted. In order to transport the spent fuel safely, a tie-down structure for supporting and transporting a cask containing the spent fuel is essential. To ensure its structural integrity, a method for finding an optimum conceptual design for the tie-down structure is presented. An optimized transportation test model of a tie-down structure for the KORAD-21 metal cask is derived based on the proposed optimization approach, and the transportation test model is manufactured by redesigning the optimized model to enable its producibility. The topology optimization approach presented in this paper can be used to obtain optimum conceptual designs of tie-down structures developed in the future.

• 기술혁신학회지
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• 제5권1호
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• pp.90-109
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• 2002

The national nuclear R&D projects have been implemented for the purpose of supply of nuclear power which was proved to be safety and stability. Evaluating the national nuclear R&D projects has critical importance in nu-clear energy management aspect. This paper employs multi-attribute utility analysis as a basis for obtaining an evaluation index to assess the national nuclear R&D projects using a specific case study of Korea. To structure and quantify basic values for the evaluation, we elicited important attributes, then refined and structured them into a hierarchy. A multi-attribute index is constructed as a multi-attribute utility function, based on value judgments provided by a group of technical experts, policy makers, and faculties. As a result, the objective of attainment of a proposed object is given the highest priority, followed by appropriateness of project and research strategy, utilization of research output, within attribute ranges defined for the attributes. We found that the work and results of this study can provide valuable insights and decision opportunities for virtually all major decision making in evaluation of the national nuclear R&D projects in Korea.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1234-1245
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• 2018

The risk of multi-unit nuclear power plants (NPPs) at a site has received considerable critical attention recently. However, current probabilistic safety assessment (PSA) procedures and computer code do not support multi-unit PSA because the traditional PSA structure is mostly used for the quantification of single-unit NPP risk. In this study, the main purpose is to develop a multi-unit Level 2 PSA method and apply it to full-power operating six-unit OPR1000. Multi-unit Level 2 PSA method consists of three steps: (1) development of single-unit Level 2 PSA; (2) extracting the mapping data from plant damage state to source term category; and (3) combining multi-unit Level 1 PSA results and mapping fractions. By applying developed multi-unit Level 2 PSA method into six-unit OPR1000, site containment failure probabilities in case of loss of ultimate heat sink, loss of off-site power, tsunami, and seismic event were quantified.

• Nuclear Engineering and Technology
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• 제46권5호
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• pp.655-666
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• 2014

The CUPID code has been developed at KAERI for a transient, three-dimensional analysis of a two-phase flow in light water nuclear reactor components. It can provide both a component-scale and a CFD-scale simulation by using a porous media or an open media model for a two-phase flow. In this paper, recent advances in the CUPID code are presented in three sections. First, the domain decomposition parallel method implemented in the CUPID code is described with the parallel efficiency test for multiple processors. Then, the coupling of CUPID-MARS via heat structure is introduced, where CUPID has been coupled with a system-scale thermal-hydraulics code, MARS, through the heat structure. The coupled code has been applied to a multi-scale thermal-hydraulic analysis of a pool mixing test. Finally, CUPID-SG is developed for analyzing two-phase flows in PWR steam generators. Physical models and validation results of CUPID-SG are discussed.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.359-359
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• 1998

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.249-249
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• 1999

• 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.