• Journal of the Korean Society of Radiology
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• v.17 no.1
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• pp.79-90
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• 2023

The study is to estimate the social costs of premature deaths due to nuclear power plant(NPP) accidents, by resorting to the contingent valuation method(CVM) which is used to estimate the value of a statistical life(VSL). The VSL estimate is about 3.55 billion won, which is multiplied by some 1.8 million premature deaths due to the accidents in world history of NPP, to get a maximum social cost of 1,952 trillion won. This estimate is equivalent to the 2022 real GDP of Korea. The annual average number of premature deaths and the resulting average social cost is 26,000 and 28 trillion won, respectively. The social cost of premature deaths due not only to accidents, but also the air pollutants from fired power plants(FPP) during 1987~2021 is estimated to be 26,919 trillion won. This is equivalent to 2021 US GDP, and is about 3,000 times higher than that for NPP of 9 trillion won. In 2021, the estimated social costs of FPP and NPP are 1,075 trillion won and 292 billion won, respectively. For South Korea, the study suggests to adapt an energy mix of increased share of electricity production for NPP relative to FPP, given that the 2050 carbon neutrality strategy of Korea is expected to lead to an increased share of renewable energy in electricity production. The study emphasizes accumulating the number of CVM-based VSL studies to ensure efficient energy policies.

• Journal of the Earthquake Engineering Society of Korea
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• v.19 no.6
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• pp.265-271
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• 2015

The tsunami hazard analysis is performed for testing the application of probabilistic tsunami hazard analysis to nuclear power plant sites in the Korean Peninsula. Tsunami hazard analysis is based on the seismic hazard analysis. Probabilistic method is adopted for considering the uncertainties caused by insufficient information of tsunamigenic fault sources. Logic tree approach is used. Uljin nuclear power plant (NPP) site is selected for this study. The tsunamigenic fault sources in the western part of Japan (East Sea) are used for this study because those are well known fault sources in the East Sea and had several records of tsunami hazards. We have performed numerical simulations of tsunami propagation for those fault sources in the previous study. Therefore we use the wave parameters obtained from the previous study. We follow the method of probabilistic tsunami hazard analysis (PTHA) suggested by the atomic energy society of Japan (AESJ). Annual exceedance probabilities for wave height level are calculated for the site by using the information about the recurrence interval, the magnitude range, the wave parameters, the truncation of lognormal distribution of wave height, and the deviation based on the difference between simulation and record. Effects of each parameters on tsunami hazard are tested by the sensitivity analysis, which shows that the recurrence interval and the deviation dominantly affects the annual exceedance probability and the wave heigh level, respectively.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1331-1344
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• 2021

This study is to evaluate the seismic capacity of the fire-damaged cabinet facility in a nuclear power plant (NPP). A prototype of an electrical cabinet is modeled using OpenSees for the numerical simulation. To capture the nonlinear behavior of the cabinet, the constitutive law of the material model under the fire environment is considered. The experimental record from the impact hammer test is extracted trough the frequency-domain decomposition (FDD) method, which is used to verify the effectiveness of the numerical model through modal assurance criteria (MAC). Assuming different temperatures, the nonlinear time history analysis is conducted using a set of fifty earthquakes and the seismic outputs are investigated by the fragility analysis. To get a threshold of intensity measure, the Monte Carlo Simulation (MCS) is adopted for uncertainty reduction purposes. Finally, a capacity estimation model has been proposed through the investigation, which will be helpful for the engineer or NPP operator to evaluate the fire-damaged cabinet strength under seismic excitation. This capacity model is presented in terms of the High Confidence of Low Probability of Failure (HCLPF) point. The results are validated by the proper judgment and can be used to analyze the influences of fire on the electrical cabinet.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.2
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• pp.91-99
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• 2016

The safety of nuclear power plants has received much attention; this safety largely depends on the continuous availability of electrical energy source during all modes of nuclear power plant operation. A station blackout (SBO) describes the loss of the off-site electric power, the failure of the emergency diesel generators, and the unavailability of the alternate AC (AAC) power. Consequently, all systems that are AC powered such as the safety injection, shutdown cooling, component cooling water, and essential service water systems are unavailable. The aim of this study is to investigate the deficiencies of the existing alternatives for coping with an extended SBO for APR1400 design. The method is analyzing the existing deficiencies and proposing an optimal solution for the NPP design during the extended SBO. This study, established a new passive system, called passive decay heat removal system (PDHRS), using systems engineering approach.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2652-2659
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• 2020

The methyl iodide (CH₃I) removal performance of tri-ethylene-di-amine impregnated activated carbon (TEDA-AC) used in the air cleaning unit of nuclear power plants (NPPs) should be maintained at least 99% between 24 month-performance test period. In order for evaluating the effectiveness of TEDA-AC on the removal performance of CH₃I in nuclear power plant during the operation of NPPs, the long-term test for up to 15 months was carried out under the simulated operating conditions (e.g., 25 ℃, RH 50%, ppb level poisoning gases injection) at nuclear power plants (NPPs). The TEDA-AC samples were analyzed with the Brunauer-Emmett-Teller (BET) specific surface area and TEDA content as well as CH₃I penetration test. It is clearly evident that more than 99% of CH₃I removal performance of TEDA-AC was observed in the TEDA-AC samples during 15 months of long-term operation under the simulated NPP operating conditions including the ppb level of organic and oxide form of poisoning gases. BET specific surface area and TEDA content that can affect the CH₃I removal performance of TEDA-AC were also maintained as those in new TEDA-AC during 15 months of long-term operation.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.899-905
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• 1998

The analysis priority makings the recommendation to reduce the total core damage frequency (CDF) of Wolsong nuclear Power Plant nits 2/3/4 was Performed in this paper. In order to derive the recommendation, the sensitivity analysis of CDF on which major contributors effect m performed based on the accident quantification results during Level 1 Probabilistic safety assessment (PSA). Priorities were ranked in tile way that compares the CDF reduction rate with efforts required to implement those recommendations using risk matrix

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2221-2229
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• 2020

After the Fukushima Daiichi nuclear power plant (NPP) accident, level 3 probabilistic safety assessment (PSA) has emerged as an important task in order to assess the risk level of the multi-unit NPPs in a single nuclear site. Accurate calculation of the radionuclide concentrations and exposure doses to the public is required if a nuclear site has multi-unit NPPs and large number of people live near NPPs. So, there has been a great need to develop a new method or procedure for the fast and accurate offsite consequence calculation for the multi-unit NPP accident analysis. Since the multi-unit level 3 PSA is being currently performed assuming that all the NPPs are located at the same position such as a center of mass (COM) or base NPP position, radionuclide concentrations or exposure doses near NPPs can be drastically distorted depending on the locations, multi-unit NPP alignment, and the wind direction. In order to overcome this disadvantage of the COM method, the idea of a new multiple location (ML) method was proposed and implemented into a new tool MURCC (multi-unit radiological consequence calculator). Furthermore, the MURCC code was further improved for the multi-unit level 3 PSA that has the arbitrary number of multi-unit NPPs. The objectives of this study are to (1) qualitatively and quantitatively compare COM and ML methods, and (2) demonstrate the strength and efficiency of the ML method. The strength of the ML method was demonstrated by the applications to the multi-unit long-term station blackout (LTSBO) accidents at the four-unit Vogtle NPPs. Thus, it is strongly recommended that this ML method be employed for the offsite consequence analysis of the multi-unit NPP accidents.

• Journal of Korean Institute of Industrial Engineers
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• v.35 no.2
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• pp.178-184
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• 2009

The Nuclear Power Plant(NPP) industry in Korea has been making efforts to reduce the human errors which have largely contributed to about 150 nuclear reactor trips since 2001. Recently, the Crew Resource Management(CRM) training has risen as an alternative countermeasure against the nuclear reactor trips caused by human errors. The effectiveness of CRM training in NPP industry, however, has not been proven to be significant yet. In this study a return on investment(ROI) model is developed to measure the effectiveness of CRM training for the operators in Korean NPP. The model consists of mathematical expressions including multiple variables affecting the CRM training impacts and nuclear reactor trips. Implication of the model is discussed further in detail.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.791-794
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• 2013

Since the Fukushima accident in 2011, the importance of the electrical systems in nuclear power plants (NPPs) has been emphasized. The result has been that NPP regulators are enhancing their monitoring of loss of offsite power (LOOP) events. Korea Hydro & Nuclear Power Co. (KHNP) is reviewing the status and issues related to LOOPs, and is attempting to establish specific countermeasures to prevent LOOPs, because they can have severe consequences in the complicated maintenance schedule during an outage. A starting point for preventing LOOPs is the control of the loss of voltage (LOV)-initiating components. In order to reflect this in the risk assessment program, an LOV monitor is being developed for use during plant outages.

• Journal of the Korea Institute of Information Security & Cryptology
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• v.24 no.3
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• pp.561-570
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• 2014

Nuclear Power Plant Instrumentation and Control System(NPP I&C) which is used to operate safely is changing from analog technology to digital technology. Ever since NPP Centrifuge of Iran Bushehr was shut down by Stuxnet attack in 2010, the possibility of cyber attacks against the NPP has been increasing. However, the domestic and international regulatory guidelines that was published to strengthen the cyber security of the NPP I&C describes security requirements and method s to establish policies and procedures. These guidelines are not appropriate for the development of real applicable cyber security technology. Therefore, specialized cyber security technologies for the NPP I&C need to be developed to enhance the security of nuclear power plants. This paper proposes a cyber security technology development system which is exclusively for the development of nuclear technology. Furthermore, this method has been applied to the ESF-CCS developed by The KINCS R&D project.