• Journal of Radiation Protection and Research
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• 제49권2호
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• pp.65-67
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• 2024

This review is based on articles published in Japanese Journal of Health Physics on "Radiation-exposed medical care and waste" related to Tokyo Electric Power Company Incorporated (TEPCO)'s Fukushima Daiichi Nuclear Power Station accident. Here, we have considered three original articles; one technical data, one special article, one 50^th anniversary article, one preface, three topics, and two cases of From Japan to the World (J to W). These articles have reported the system and standards that were established after the accident. Moreover, they have summarized rare experiences such as the fumbling response at the time and the evaluation of samples in that disaster. These articles constitute valuable records of the situation.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 추계학술발표대회 요약집
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• pp.186-186
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• 2003

• 한국환경보건학회지
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• 제37권3호
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• pp.226-233
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• 2011

Exactly 25 years after the Chernobyl nuclear disaster, 11th of March 2011 the Fukushima nuclear accident occurred in Japan and was ranked at level 7 same to the Chernobyl. A Korean and Japanese joint civil survey was conducted around Fukushima on April 13-17. The radiation survey data clearly shows a large hotspot area between 20 km and 50 km radius north and west direction from the accident reactors, with the highest radiation recorded being 55.64 ${\mu}Sv/hr$ in the air, 99.89 ${\mu}Sv/hr$ in the surface air, and 36.16 ${\mu}Sv/hr$ in a car, respectively. 3.65 ${\mu}Sv/hr$ in the air and 6.89 ${\mu}Sv/hr$ in the surface air were detected at the playground of an elementary school in Fukushima City. Spring came with full cherry blossoms in Fukushima, but it was silent spring of radiation contamination. Interviews with Fukushima nuclear refugees reveal serious problems about Japanese nuclear safety systems, such as there was no practical evacuation drill within 1-10 km and no plan at all for 10-30 km areas. Several reforms items for Korean nuclear safety system can be suggested: minimization of accident damage, clear separation of regulatory and safety bureaus with a new and independent administrating agency, community participation and agreement regarding the safety system and levels, which is the major concern of 80% Korean. To tackle threats of nuclear disaster in neighboring nations like China, a new position entitled 'Ambassador for nuclear safety diplomacy' is highly necessary. The nuclear safety of Korea should no longer be the monopoly of those nuclear engineers and limited technocrats criticized as a 'nuclear mafia'.

• STI Policy Review
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• 제4권1호
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• pp.135-161
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• 2013

This paper examines Japan's Science and Technology (S&T) Basic Plans in accordance with its S&T Basic Law. The Basic Plans promote two major innovation (Green Innovation and Life Innovation) towards the creation of new markets and jobs, specifically under the Fourth S&T Basic Plan enacted on August 2011. Successful smart community demonstration projects at four urban localities were launched under plans to promote Green Innovation research and development of renewable energy technologies. However, the expectation that renewable energy such as solar or wind power can replace nuclear power is not backed by sufficient evidence. Furthermore, the electricity produced by these sources is expensive and unstable owing to its reliance on weather conditions. The Fukushima nuclear power plant accident on March 2011 has also seriously affected Japan's future energy plans. According to a government estimate, electricity charges would double if nuclear power generation were abandoned, imposing a heavy burden on the Japanese economy. Japan is in need of energy policies designed on the basis of more far-sighted initiatives.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4146-4158
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• 2023

This paper proposes a combined plastic and creep constitutive model of A533B1 pressure vessel steel to simulate progressive deformation of nuclear pressure vessels under severe accident conditions. To develop the model, recent tensile test data covering a wide range of temperatures (from RT to 1,100 ℃) and strain rates (from 0.001%/s to 1.0%/s) was used. Comparison with experimental data confirms that the proposed combined plastic and creep model can well reflect effects of temperature and strain rate on tensile behaviour up to failure. In the companion paper (Part II), the proposed model will be used to simulate OECD lower head failure (OLHF) test data.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.33-46
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• 2015

During the past 10 years, the Korea Atomic Energy Research Institute (KAERI) has performed a study to control hydrogen gas in the containment of the nuclear power plants. Before the Fukushima accident, analytical activities for gas distribution analysis in experiments and plants were primarily conducted using a multidimensional code: the GASFLOW. After the Fukushima accident, the COM3D code, which can simulate a multidimensional hydrogen explosion, was introduced in 2013 to complete the multidimensional hydrogen analysis system. The code validation efforts of the multidimensional codes of the GASFLOW and the COM3D have continued to increase confidence in the use of codes using several international experimental data. The OpenFOAM has been preliminarily evaluated for APR1400 containment, based on experience from coded validation and the analysis of hydrogen distribution and explosion using the multidimensional codes, the GASFLOW and the COM3D. Hydrogen safety in nuclear power has become a much more important issue after the Fukushima event in which hydrogen explosions occurred. The KAERI is preparing a large-scale test that can be used to validate the performance of domestic passive autocatalytic recombiners (PARs) and can provide data for the validation of the severe accident code being developed in Korea.

• 전기학회논문지
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• 제65권5호
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• pp.836-842
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• 2016

When an accident occurs in the nuclear power plant, the faulted information might mislead to the high possibility of aggravating the accident. At the Fukushima accident, the operators misunderstood that there was no core exposure despite in the processing of core damage, because the instrument information of the reactor water level was provided to the operators optimistically other than the actual situation. Thus, this misunderstanding actually caused to much confusions on the rapid countermeasure on the accident, and then resulted in multiplying the accident propagation. It is necessary to be equipped with the function that informs operators the status of instrument integrity in real time. If plant operators verify that the instruments are working properly during accident conditions, they are able to make a decision more safely. In this study, we have performed various tests for the fault detection sensitivity of an data-driven empirical model to review the usability of the model in the accident conditions. The test was performed by using simulation data from the compact nuclear simulator that is numerically simulated to PWR type nuclear power plant. As a result of the test, the proposed model has shown good performance for detecting the specified instrument faults during normal plant conditions. Although the instrument fault detection sensitivity during plant accident conditions is lower than that during normal condition, the data-drive empirical model can be detected an instrument fault during early stage of plant accidents.

• Nuclear Engineering and Technology
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• 제41권1호
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• pp.63-70
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• 2009

A steam generator tube rupture (SGTR) accident, which is a partial reactor building bypass scenario, has a low probability and high consequences. SAMG has been used to manage the progression of severe accidents and the release of fission products induced by an SGTR at the Wolsong plants. Four of the six SAGs in the SAMG are used to manage the progression of a severe accident induced by an SGTR at the Wolsong plants. The results of the ISAAC code calculation have shown that the proper use the SAMG can stop a severe accident from progressing and keep the reactor building intact during a severe accident. These results confirm that the SAMG is an effective means of managing the progression of severe accidents initiated by an SGTR at the Wolsong plants.

• 산업공학
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• 제15권2호
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• pp.114-125
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• 2002

The objective of this paper is to describe a systematic accident scenario analysis method(SASA) adept at creating accident scenarios for the design of safer products. This approach was inspired by the Quality Function Deployment(QFD) method, which is conventionally used in quality management. In this study, the QFD provides a formal and systematic scheme to devise accident scenarios while maintaining objectivity. SASA consists of three key stages to be broken down into a series of consecutive steps:(1) developing an accident analysis tableau,(2) devising the accident scenarios using the accident analysis tableau,(3) performing a feasibility test, a clustering process and a patterning process, and finally(4) performing quantitative evaluation of each accident scenario. The SASA was applied to a case study of child safety seats. The accident analysis tableau devised 2828(maximum) accident scenarios from all possible relationships between the hazard factors and situation characteristics. Among them, 270 scenarios were devised through the feasibility test and the clustering process. The patterning process reduced them to 29 patterns representative of all accident scenarios. Based on an intensive analysis of the accident patterns, design guidelines for a safer child safety seat were recommended. The implications of the study on the child safety seat case were then discussed.

• 시스템엔지니어링학술지
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• 제19권2호
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• pp.46-58
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• 2023

Deterministic safety analysis is a crucial part of safety assessment, particularly when it comes to demonstrating the safety of nuclear power plant designs. The traditional approach to deterministic safety analysis models is to model the nuclear core using point kinetics. However, this simplified approach does not fully reflect the real core behavior with proper moderator and fuel reactivity feedbacks during the transient. The use of Multi-Physics approach allows more precise simulation reflecting the inherent three-dimensionality (3D) of the problem by representing the detailed 3D core, with instantaneous updates of feedback mechanisms due to changes of important reactivity parameters like fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC). This paper addresses a CEA ejection accident at hot full power (HFP), in which the underlying strong and un-symmetric feedback between thermal-hydraulics and reactor kinetics exist. For this purpose, a multi-physics analysis tool has been selected with the nodal kinetics code, 3DKIN, implicitly coupled to the thermal-hydraulic code, RELAP5, for real-time communication and data exchange. This coupled approach enables high fidelity three-dimensional simulation and is therefore especially relevant to reactivity initiated accident (RIA) scenarios and power distribution anomalies with strong feedback mechanisms and/or un-symmetrical characteristics as in the CEA ejection accident. The Systems Engineering approach is employed to provide guidance in developing the work in a systematic and efficient fashion.