• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1830-1837
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• 2023

The thermal cutting of contaminated or activated metals during decommissioning nuclear power plants inevitably results in the release of radioactive aerosol. Since radioactive aerosols are pernicious particles that contribute to the internal dose of workers, air conditioning units with a HEPA filter are used to remove radioactive aerosols. However, a HEPA filter cannot be used permanently. This study evaluates the efficiency and lifetime of filters in actual metal cutting condition using a plasma arc cutter and a high-resolution aerosol detector. The number concentration and size distribution of aerosols from 6 nm to 10 ㎛ were measured on both the upstream and downstream sides of the filter. The total aerosol removal efficiency of HEPA filter satisfies the standard of removing at least 99.97% of 0.3 ㎛ airborne particles, even if the pressure drop increases due to dust feeding load. The pressure drop and particle size removal efficiency at 0.3 ㎛ of the HEPA filter were found to increase with repeated cutting experiments. By contrast, the efficiency of used HEPA filter reduced in removing nano-sized aerosols by up to 79.26%. Altogether, these results can be used to determine the performance guidance and replacement frequency of HEPA filters used in nuclear power plants.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4181-4190
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• 2023

Nuclear power plants are subjected to various processes during decommissioning, including cutting, decontamination, disposal, and treatment. The cutting of massive bio-shields is a significant step in the decommissioning process. Cutting is performed near the target structure, and during this process, workers are exposed to potential radioactive elements. However, studies considering worker exposure management during such cutting operations are limited. Furthermore, dismantling a nuclear power plant under certain circumstances may result in the unnecessary radiation exposure of workers and an increase in secondary waste generation. In this study, a cutting scenario was formulated considering the bio-shield as a representative structure. The specifications of a standard South Korean radioactive waste disposal drum were used as the basic conditions. Additionally, we explored the hot-to-cold and cold-to-hot methods, with and without the application of polishing during decontamination. For evaluating various scenarios, different cutting time points up to 30 years after permanent shutdown were considered, and cutting speeds of 1-10nullm²/h were applied to account for the variability and uncertainty attributable to the design output and specifications. The obtained results provide fundamental guidelines for establishing cutting methods suitable for large structures.

• Journal of the Korean Society of Safety
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• v.27 no.5
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• pp.224-228
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• 2012

According to the results of Probabilistic Safety Assessment(PSA) for a Nuclear Power Plant(NPP), an Emergency Power Supply(EPS) system has been considered as one of the most important safety system. Especially, the interests in the reliability of the EPS system have been increased after the severe accidents of Fukushima Daiichi. Firstly, we performed the risk assessment and the importance analysis of the EPS system based on the PSA models of the reference plant, which is the Korean standard NPP type. Considering a portable Diesel Generator(DG) system as the reliability reinforcement of the EPS system, we modified the PSA models and performed the risk impact assessment and the importance analysis. Although the reliability of the potable DG could be about 20% of the reliability of the alternative AC DG, we identified that Core Damage Frequency(CDF) was decreased by at least 4.6%. In addition, the risk impacts due to the unavailability of the EPS system on CDF were decreased.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.569-580
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• 2014

Seismic isolation is a viable strategy for protecting safety-related nuclear structures from the effects of moderate to severe earthquake shaking. Although seismic isolation has been deployed in nuclear structures in France and South Africa, it has not seen widespread use because of limited new build nuclear construction in the past 30 years and a lack of guidelines, codes and standards for the analysis, design and construction of isolation systems specific to nuclear structures. The funding by the United States Nuclear Regulatory Commission of a research project to the Lawrence Berkeley National Laboratory and MCEER/University at Buffalo facilitated the writing of a soon-to-be-published NUREG on seismic isolation. Funding of MCEER by the National Science Foundation led to research products that provide the technical basis for a new section in ASCE Standard 4 on the seismic isolation of safety-related nuclear facilities. The performance expectations identified in the NUREG and ASCE 4 for seismic isolation systems, and superstructures and substructures are described in the paper. Robust numerical models capable of capturing isolator behaviors under extreme loadings, which have been verified and validated following ASME protocols, and implemented in the open source code OpenSees, are introduced.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.1
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• pp.1-10
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• 2022

Currently, there is no technical standard and regulation for seismic analysis of non-nuclear safety piping. Accordingly, ASME Sec.III ND, a standards applied to safety class 3 piping, is applied. However, the technical standard applied for other than seismic analysis is ASME B31, which leads to controversy. In this study, the feasibility of applying ASME B31E was confirmed by reviewing rulescomparing technical standards, and evaluating piping allowable stress margins. The evaluation revealed that applying ASME B31.1 as a technical standard is too conservative compared to ASME Sec.III ND. On the other hand, ASME B31E (issued at the request of the industry) clearly presents the technical standards for seismic analysis of ASME B31 piping, and shows a similar level of conservatism compared to ASME Sec.III ND. It is expected to reduce the controversy over technical standards for seismic analysis of non-nuclear safety piping by applying ASME B31E.

• The Transactions of the Korean Institute of Power Electronics
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• v.18 no.4
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• pp.397-402
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• 2013

This paper proposes a new passive type LED driver which satisfies the standard of power factor (PF) and total harmonic distortion (THD). The proposed passive LED driver also has high-efficiency and long-life time characteristics compared to active LED driver which is composed of op-amp, switches and so on. By using just passive components such as inductor, capacitor, and diode, it has resolved extremely short-life time and low-efficiency problems of previous LED drivers. It has achieved PF of 0.99, THD of 16.4 %, and the total efficiency of 95 %. The proposed passive LED driver is fully analyzed and verified by simulations and experiments, which results are in good agreement each other.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.2
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• pp.66-70
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• 2016

Following the 2011 Fukushima Nuclear Power Plant accident, the IAEA has issued a revised version of the Nuclear Safety Standard for beyond design basis earthquake to consider the core meltdown accident. In Korea, relevant laws and regulations were also revised to consider beyond design basis earthquake to nuclear components. In this paper, CAV, an seismic damage factor that determines the restart of nuclear power plant after operating breakdown earthquake, is proposed for extension to the beyond design basis earthquake. For pipings not satisfying the beyond design basis earthquake condition, several evaluation methods are suggested, such as strain-based evaluation methods, simple nonlinear analysis method and cumulative damage evaluation method.

• Journal of the Korean Society of Safety
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• v.25 no.6
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• pp.53-59
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• 2010

A fatigue damage caused by cyclic load is considered as one of the important failure mechanisms that threaten the integrity of structures and components in a nuclear power plant. In ASME code section III NB, the fatigue analysis procedure and standard S-N curves for the class 1 components are described and these criteria should be met at the design step of components. As the current ASME S-N curves are based on the very conservative assumptions such as a local stress concentration effect, immoderate transient frequencies and a constant Young's modulus, however, they can not precisely address the fatigue behavior of components. In order to find out the technical solution for these problems, a number of researches and discussion have been carried out continuously at home and abroad over the decades. In this study, detailed fatigue analyses for DVI nozzle with various mesh density of finite elements were performed to evaluate effect of stress concentration factors on the fatigue analysis procedure and the excessive conservatism of stress concentration factors are confirmed through the analysis results.

• Nuclear Engineering and Technology
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• v.36 no.2
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• pp.184-195
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• 2004

The multi-tasking feature of digital I&C equipment could increase risk concentration because the I&C equipment affects the actuation of the safety functions in several ways. Anticipated Transient without Scram (ATWS) is a typical case of safety function failure in nuclear power plants. In a conventional analysis, mechanical failures are treated as the main contributors of the ATWS. This paper quantitatively presents the probability of the ATWS based on a fault tree analysis of a Korea Standard Nuclear Power Plant is also presented. An analysis of the digital equipment in the digital plant protection system. The results show that the digital system severely affects the ATWS frequency. We also present the results of a sensitivity study, which show the effects of the important factors, and discuss the dependency between human operator failure and digital equipment failure.

• The Journal of the Korea institute of electronic communication sciences
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• v.10 no.1
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• pp.81-94
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• 2015

The purpose of this study is to find ways to build a desirable risk governance to respond to these uncertainties like such as increased risk due to the rapid development of modern science and technology in S. Korea. This study is about model of risk communication with scientific technology for additional construction of nuclear power plants. This study analyzed risk communication with scientific technology through new explanatory models between approval opinions of scientific professional group and opposition opinions of the public with Kori nuclear power plant of S. Korea. And then This study investigated related system for policy promotion in the field of nuclear energy of the current S. Korea. Consequently, Governance to strengthen the negotiations on nuclear technology at the time of the interaction in risk communication have been identified and the future of policy direction of the Korean Standard Nuclear Power Plant was detected.