• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1297-1306
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• 2019

Probabilistic safety assessments (PSA) have been used for several decades to visualize the risk level of commercial nuclear power plants (NPPs). Since the role of a human reliability analysis (HRA) is to provide human error probabilities for safety critical tasks to support PSA, PSA quality is strongly affected by HRA quality. Therefore, it is important to understand the underlying limitations or problems of HRA techniques. For this reason, this study conducted a survey among 14 subject matter experts who represent the HRA community of domestic Korean NPPs. As a result, five significant HRA issues were identified: (1) providing a technical basis for the K-HRA (Korean HRA) method, and developing dedicated HRA methods applicable to (2) diverse external events to support Level 1 PSA, (3) digital environments, (4) mobile equipment, and (5) severe accident management guideline tasks to support Level 2 PSA. In addition, an HRA method to support multi-unit PSA was emphasized because it plays an important role in the evaluation of site risk, which is one of the hottest current issues. It is believed that creating such a catalog of prioritized issues will be a good indication of research direction to improve HRA and therefore PSA quality.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.967-973
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• 2021

In 2011, an earthquake and subsequent tsunami hit the Fukushima Daiichi Nuclear Power Plant, causing simultaneous accidents in several reactors. This accident shows us that if there are several reactors on site, the seismic risk to multiple units is important to consider, in addition to that to single units in isolation. When a seismic event occurs, a seismic-failure correlation exists between the nuclear power plant's structures, systems, and components (SSCs) due to their seismic-response and seismic-capacity correlations. Therefore, it is necessary to evaluate the multi-unit seismic risk by considering the SSCs' seismic-failure-correlation effect. In this study, a methodology is proposed to obtain the seismic-response-correlation coefficient between SSCs to calculate the risk to multi-unit facilities. This coefficient is calculated from a probabilistic multi-unit seismic-response analysis. The seismic-response and seismic-failure-correlation coefficients of the emergency diesel generators installed within the units are successfully derived via the proposed method. In addition, the distribution of the seismic-response-correlation coefficient was observed as a function of the distance between SSCs of various dynamic characteristics. It is demonstrated that the proposed methodology can reasonably derive the seismic-response-correlation coefficient between SSCs, which is the input data for multi-unit seismic probabilistic safety assessment.

• 한국안전학회지
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• 제39권1호
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• pp.96-103
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• 2024

The demand for mass calculation of offsite consequence analysis to conduct exhaustive single-unit or multi-unit Level 3 PSA is increasing. In order to perform efficient offsite consequence analyses, the Korea Atomic Energy Research Institute is conducting model optimization studies to minimize the analysis time while maintaining the accuracy of the results. A previous study developed a model optimization method using efficient plume segmentation and verified its effectiveness. In this study, we investigated the possibility of optimizing the model through particle size distribution setting by checking the reduction in analysis time and deviation of the results. Our findings indicate that particle size distribution setting affects the results, but its effect on analysis time is insignificant. Therefore, it is advantageous to set the particle size distribution as fine as possible. Furthermore, we evaluated the effect of multithreading and confirmed its efficiency. Future optimization studies should be conducted on various input factors of offsite consequence analysis, such as spatial grid settings.

• Nuclear Engineering and Technology
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• 제56권9호
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• pp.3851-3863
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• 2024

The speed of offsite consequence analysis is highly important due to the extensive calculations required to handle all the scenarios for a single-unit or multi-unit Level 3 PSA (probabilistic safety assessment). To perform an offsite consequence analysis as part of Level 3 PSA, various input parameters are considered, amongst which certain parameters, such as plume segments, spatial grids, and particle size distributions, have flexible input formats. This study describes the development of an effective optimization method to reduce the analysis time as much as possible while maintaining the accuracy of the offsite consequence analysis results. The effect of plume segmentation on offsite consequence analysis was investigated by observing deviations in analysis results and changes in the required analysis times following changes in plume release. Then a plume segmentation optimization method based on the cumulative release fraction slope was developed to intensively analyze the sections with rapid release and to simplify the analysis for the sections with nonsignificant release. As a result of applying this method, the analysis time was reduced by about 54.5 % compared to the base case, while the resulting health effects showed very small deviations of 0.03 % and 1.77 % for early fatality risk and cancer fatality risk, respectively.

• 방사성폐기물학회지
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• 제14권4호
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• pp.423-434
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• 2016

인접 국가인 일본의 후쿠시마 원전에서 극한 자연재해로 인한 중대사고가 발생하면서, 국내에서 중대사고 및 확률론적 안전성 평가 (PSA, Probabilistic Safety Assessment)에 대한 중요성이 재인식되었다. 국내에서는 원전의 소외결말을 평가하는 3단계 PSA에 대한 연구개발이 최근까지 거의 이루어지지 않았다. 본 논문에서는 국외 3단계 PSA 전산코드 중, 미국의 MACCS2 (MELCORE Accident Consequence Code System 2), 유럽의 COSYMA (COde SYstem from Maria) 그리고 일본의 OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents)에 대한 간략한 분석과 미국의 MACCS2에 대한 단점 및 한계점 분석을 수행하였다. 국내 외 전문가들에 의해 공통적으로 지적되어 온 MACCS2의 한계점은 다수호기사고와 사용후핵연료 저장조로부터의 방출 모사의 불가능, 그리고 대기확산모델을 단순 가우시안 플륨모델을 기본으로 한다는 것이며, 이중 일부는 MACCS2업데이트 버전을 통해 개선되어 왔다. Food chain 모델의 모사의 제한, 해양 및 수계 확산모델의 부재, 제한된 범위의 경제영향평가 등 또한 개선되어야 할 사항이다. 기술보고의 결과는 국내 3단계 PSA 관련 기술 개발을 위한 기초자료로 활용될 수 있을 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1225-1233
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• 2024

Level 3 probabilistic safety assessment (PSA) is performed to calculate radionuclide concentrations and exposure dose resulting from nuclear power plant accidents. To calculate the external exposure dose from the released radioactive materials, the radionuclide concentrations are multiplied by two factors of dose coefficient and a finite cloud dose correction factor (FCDCF), and the obtained values are summed. This indicates that a standard set of FCDCFs is required for external exposure dose calculations. To calculate a standard set of FCDCFs, the effective distance from the release point to the receptor along the wind direction should be predetermined. The TID-24190 document published in 1968 provides equations to calculate FCDCFs and the resultant standard set of FCDCFs. However, it does not provide any explanation on the effective distance required to calculate the standard set of FCDCFs. In 2021, Sandia National Laboratories (SNLs) proposed a method to predetermine finite effective distances depending on the atmospheric stability classes A to F, which results in six standard sets of FCDCFs. Meanwhile, independently of the SNLs, the authors of this paper discovered that an infinite effective distance assumption is a very reasonable approach to calculate one standard set of FCDCFs, and they implemented it into the multi-unit radiological consequence calculator (MURCC) code, which is a post-processor of the level 3 PSA codes. This paper calculates and compares short- and long-range FCDCFs calculated using the TID-24190, SNLs method, and MURCC method, and explains the strength of the MURCC method over the SNLs method. Although six standard sets of FCDCFs are required by the SNLs method, one standard sets of FCDCFs are sufficient by the MURCC method. Additionally, the use of the MURCC method and its resultant FCDCFs for level 3 PSA was strongly recommended.