• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2018년도 춘계 학술논문 발표대회
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• pp.151-152
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• 2018

In the domestic nuclear power industry, concern about safety of nuclear power plants is continuously increased with the Fukushima nuclear power plant accident. In order to enhance the safety of nuclear power plants, it is important to ensure that the power plants are operating with proper margin within the original design bases. Margin management is the process of ensuring that the NPP designer and operator are aware of the physical and operating limits, and potential and probability of failure, for each component in the plant. All components are subject to margin considerations, but the most important components by scope and attention are those related to safety-related systems and NPP safe shutdown.

• 과학교육연구지
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• 제37권2호
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• pp.294-309
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• 2013

이 연구에서는 초등 예비교사들의 과학과 관련된 사회적 문제에 대한 인식을 알아보았다. 연구를 위해 SSI와 관련된 구체적인 이슈로 후쿠시마 원전 사고를 이용하였다. 연구에는 교원양성대학에 재학중인 대학교 2학년생 12명이 참여하였으며, 연구 참여자 모두는 초등학교 과학 교육과정에 대한 내용으로 구성되어 있는 초등 과학교육론을 수강하고 있는 학생들이었다. 연구결과 초등 예비교사들은 모두 후쿠시마 원전 사고에 대해 인지하고 있었으며, 원전 사고에 대해 다양한 매체를 통해 접하고 있음을 알 수 있었다. 특히 인터넷을 많이 사용하는 대학생이어서인지 다른 매체에 비해 인터넷을 통해 원전 사고에 대한 내용을 많이 접하고 있었으며, 또한 인터넷을 통해 추가적인 정보를 얻기도 하였다. 그러나 초등 예비교사는 원전 사고에 대해 인지하고 있는 것과는 달리 많은 정보를 갖고 있지는 않았으며, 원전 사고와 같은 SSI에 대해 평소 관심을 갖고 있거나 교육을 받은 경험을 미비했다. 원전 사고와 같은 SSI에 대해 관심은 없으나 원전 사고와 같은 과학적인 문제들이 사회에 밀접한 영향을 미친다는 것은 인식하고 있는 것으로 나타났다. 아울러 현행 교육과정에 SSI 교육이 적용되어야 한다는 점에서도 모두 공감하고 있었으며, 적용되지 않고 있다는 점에 대해서도 문제인식을 같이 하였다. 이상의 연구 결과를 보면, 과학교육의 목표 중에 하나인 과학의 본성을 이해시키는 측면과 과학 교과에 대한 흥미 제고를 위해서라도 SSI 교육에 더욱 활발한 연구뿐만 아니라 정규 교육과정에의 적용은 매우 중요하며 시급하다고 할 수 있다.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2960-2973
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• 2022

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium-concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

• 방사선산업학회지
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• 제9권4호
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• pp.209-216
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• 2015

It has been about 5 years since the Fukushima nuclear power plant accident, which contaminated large area with radioactive materials. It is necessary to assess radiation dose to establish evacuation areas and to set decontamination goal for the large contaminated area. In this study, we assessed temporal trend of radiation dose to the public living in the large area contaminated with radioactive materials after the Fukushima nuclear power plant accident. The dose assessment was performed based on Chernobyl model and RESRAD model for two evacuation lift areas, Kawauchi and Naraha. It was reported that deposition densities in the areas were $4.3{\sim}96kBq\;m^{-2}$ for $^{134}Cs$, $1.4{\sim}300kBq\;m^{-2}$ for $^{137}Cs$, respectively. Radiation dose to the residents depended on radioactive cesium concentrations in the soil, ranging $0.11{\sim}2.4mSv\;y^{-1}$ at Kawauchi area and $0.69{\sim}1.1mSv\;y^{-1}$ at Naraha area in July 2014. The difference was less than 5% in radiation doses estimated by two different models. Radiation dose decreased with calendar time and the decreasing slope varied depending on dose assessment models. Based on the Chernobyl dosimetry model, radiation doses decreased with calendar time to about 65% level of the radiation dose in 2014 after 1 year, 11% level after 10 years, and 5.6% level after 30 years. RESRAD dosimetry model more slowly decreased radiation dose with time to about 85% level after 1 year, 40% level after 10 years, and 15% level after 30 years. The decrease of radiation dose can be mainly attributed into radioactive decays and environmental transport of the radioactive cesium. Only environmental transports of radioactive cesium without consideration of radioactive decays decreased radiation dose additionally 43% after 1 year, 72% after 3 years, 80% after 10 years, and 83% after 30 years. Radiation doses estimated with cesium concentration in the soil based on Chernobyl dosimetry model were compared with directly measured radiation doses. The estimated doses well agreed with the measurement data. This study results can be applied to radiation dose assessments at the contaminated area for radiation safety assurance or emergency preparedness.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.376-385
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• 2021

Probabilistic risk assessment (PRA) has improved its elemental technologies used for assessing external events since the Fukushima Daiichi Nuclear Power Station Accident in 2011. HRA needs to be improved for analyzing tasks performed under extreme conditions (e.g., different actors responding to external events or performing operations using portable mitigation equipment). To make these improvements, it is essential to understand plant-specific and scenario-specific conditions that affect human performance. The Nuclear Risk Research Center (NRRC) of the Central Research Institute of Electric Power Industry (CRIEPI) has developed an HRA guide that compiles qualitative analysis methods for collecting plant-specific and scenario-specific conditions that affect human performance into "narratives," reflecting the latest research trends, and models for analysis of tasks under extreme conditions.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.625-636
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• 2013

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

• Nuclear Engineering and Technology
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• 제47권2호
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• pp.157-164
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• 2015

Nuclear power plants have logical loop structures in their system configuration. This paper explains the method to solve a loop structure in reliability analysis. As examples of loop structured systems, the reactor core isolation cooling system and high-pressure core injection system of a boiling water reactor are considered and analyzed under a station blackout accident condition. The analysis results show the important role of loop structures under severe accidents. For the evaluation of the safety of nuclear power plants, it is necessary to accurately evaluate a loop structure's reliability.

• Journal of Radiation Protection and Research
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• 제48권3호
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• pp.131-143
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• 2023

Background: This study examined the detection limit of thyroid screening monitoring conducted at the time of the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident in 2011 using a Monte Carlo simulation. Materials and Methods: We calculated the detection limit of a NaI(Tl) survey meter to measure ¹³¹I accumulation in the thyroid gland of children. Mathematical phantoms of 1- and 5-year-old children were developed in the simulation of the Particle and Heavy Ion Transport code System code. Contamination of the body surface with eight radionuclides found after the FDNPP accident was assumed to have been deposited on the neck and shoulder area. Results and Discussion: The detection limit was calculated as a function of ambient dose rate. In the case of 40 Bq/cm² contamination on the body surface of the neck, the present simulations showed that residual thyroid radioactivity corresponding to thyroid dose of 100 mSv can be detected within 21 days after intake at the ambient dose rate of 0.2 µSv/hr and within 11 days in the case of 2.0 µSv/hr. When a time constant of 10 seconds was used at the dose rate of 0.2 µSv/hr, the estimated survey meter output error was 5%. Evaluation of the effect of individual differences in the location of the thyroid gland confirmed that the measured value would decrease by approximately 6% for a height difference of ±1 cm and increase by approximately 65% for a depth of 1 cm. Conclusion: In the event of a nuclear disaster, simple measurements carried out using a NaI(Tl) scintillation survey meter remain effective for assessing ¹³¹I intake. However, it should be noted that the presence of short-half-life radioactive materials on the body surface affects the detection limit.

• 한국수산과학회지
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• 제49권2호
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• pp.234-240
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• 2016

We examined the diffusion of contaminants released from the southern coast around Fukushima, Japan, during the passage of typhoons using a three-dimensional numerical model (POM) to track diffusing radioactivity (RA) released from the nuclear power plant at Fukushima following the accident caused by the giant tsunami event in March 2011. Radioactive contaminants released during the passage of typhoons may have significantly affected not only Japanese but also Korean coastal waters. The model domain covered most of the northwestern Pacific including marginal seas such as the East/Japan Sea and the Yellow Sea. Several numerical experiments were conducted case studies focusing on the westward diffusion from the southern coast of Japan of contaminants derived from the source site (Fukushima) according to various attributes of the typhoons, such as intensity, track, etc. The model produced the following results 1) significant amounts of contaminants were transported in a westward direction by easterly winds favorable for generating a coastal air stream along the southern Japanese coast, 2) the contaminants reached as far as Osaka Bay with the passage of typhoons, forced by a 5-day positive sinusoidal form with a (right-) northward track east of Fukushima, and 3) the range of contamination was significant, extending to the interior of the East/Japan Sea around the Tsugaru Strait. The model suggests that contaminants and/or radioactivity released from Fukushima with the passage of typhoons can affect Korean waters including the northeastern East/Japan Sea around the Tsugaru Strait, especially when the typhoon tracks are favorable for generating a westward coastal air stream along the southern Japanese coast.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.597-604
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• 2013

Containment Filtered Vent Systems (CFVSs) have been mainly equipped in nuclear power plants in Europe and Canada for the controlled depressurization of the containment atmosphere under severe accident conditions. This is to keep the containment integrity against overpressure during the course of a severe accident, in which the radioactive gas-steam mixture from the containment is discharged into a system designed to remove the radionuclides. In Korea, a CFVS was first introduced in the Wolsong unit-1 nuclear power plant as a mitigation measure to deal with the threat of over pressurization, following post-Fukushima action items. In this paper, the overall features of a CFVS installation such as risk assessments, an evaluation of the performance requirements, and a determination of the optimal operating strategies are analyzed for the Wolsong unit 1 nuclear power plant using a severe accident analysis computer code, ISAAC.