• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.225-236
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• 2012

The Fukushima-Daiichi accident shook the world, as a well-known plant design, the General Electric BWR Mark I, was heavily damaged in the tsunami, which followed the Great Japanese Earthquake of 11 March 2011. Plant safety functions were lost and, as both AC and DC failed, manoeuvrability of the plants at the site virtually came to a full stop. The traditional system of Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMG) failed to protect core and containment, and severe core damage resulted, followed by devastating hydrogen explosions and, finally, considerable radioactive releases. The root cause may not only have been that the design against tsunamis was incorrect, but that the defence against accidents in most power plants is based on traditional assumptions, such as Large Break LOCA as the limiting event, whereas there is no engineered design against severe accidents in most plants. Accidents beyond the licensed design basis have hardly been considered in the various designs, and if they were included, they often were not classified for their safety role, as most system safety classifications considered only design basis accidents. It is, hence, time to again consider the Design Basis Accident, and ask ourselves whether the time has not come to consider engineered safety functions to mitigate core damage accidents. Associated is a proper classification of those systems that do the job. Also associated are safety criteria, which so far are only related to 'public health and safety'; in reality, nuclear accidents cause few casualties, but create immense economical and societal effects-for which there are no criteria to be met. Severe accidents create an environment far surpassing the imagination of those who developed EOPs and SAMG, most of which was developed after Three Mile Island - an accident where all was still in place, except the insight in the event was lost. It requires fundamental changes in our present safety approach and safety thinking and, hence, also in our EOPs and SAMG, in order to prevent future 'Fukushimas'.

• Journal of Advanced Marine Engineering and Technology
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• 제40권3호
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• pp.165-173
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• 2016

2011년 3월 11일 리히터 스케일 9.0의 강진과 10-14m파도로 인해 Fukushima Daiichi(FD) 원자력 단지의 주전력과 보조전력이 끊어져 냉각장치가 작동하지 않았고 노심의 열이 제거되지 못해 폭발이 일어나는 사고가 발생했다. 노심의 열이 제거되지 못하면 핵연료 피복재인 지르칼로이(zircaloy)와 같은 금속이 고온 상태에서 수증기와 산화 반응하여 수소를 발생시킨다. 발생된 수소는 격납건물로 방출되는데 방출된 수소가 연소하는 경우 격납건물의 안정성에 영향을 줄 정도의 큰 충격을 유발할 수 있는 수소폭발로 이어질 수 있다. 본 연구에서는 격납건물 내부의 수소 분포를 분석한 연구 [1]에서 제시한 폭발의 위해도가 높은 영역에 대하여 폭발해석을 수행하였으며 수소 폭발이 격납건물의 건전성에 미치는 영향에 대하여 분석하였다. 격납건물 중앙부를 제외하고 수소폭발이 발생하였고 상부에 전체 수소의 40%이상이 모였을 때와 하부 좌측, 우측의 격벽사이에 수소가 모였을 때 큰 폭발이 발생했으며 격납건물 벽면에 큰 응력을 동반하였다.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1246-1254
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• 2018

The importance of performing Level 3 probabilistic safety assessments (PSA) along with a general interest in assessing multi-unit risk has been sharply increasing after the Fukushima Daiichi nuclear power plant (NPP) accident. However, relatively few studies on multi-unit Level 3 PSA have been performed to date, reflecting limited scenarios of multi-unit accidents with higher priority. The major difficulty to carry out a multi-unit Level 3 PSA lies in the exponentially increasing number of multi-unit accident combinations, as different source terms can be released from each NPP unit; indeed, building consequence models for the astronomical number of accident scenarios is simply impractical. In this study, a new approach has been developed that employs the look-up table method to cover every multi-unit accident scenario. Consequence results for each scenario can be found on the table, established with a practical amount of effort, and can be matched to the frequency of the scenario. Preliminary application to a six-unit NPP site was carried out, where it was found that the difference between full-coverage and cut-off cases could be considerably high and therefore influence the total risk. Additional studies should be performed to fine tune the details and overcome the limitations of the approach.

• 신재생에너지
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• 제7권4호
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• pp.37-41
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• 2011

This paper presents a Combined solar-wave Power Generation (CPG) concept that the CPG unit is maintained as buoyant at the level of sea water and it is also supported by a submerged tunnel, with the aim of supplying emergency electric power during the station blackout events of nuclear power plants. The CPG concept has been motivated from the 2011 Fukushima-Daiichi Accidents due to the loss of both offsite AC power and emergency diesel power caused by natural hazards such as earthquake and tsunami. The CPG is conceptualized by applying different types and different sites for emergency power generation, in order to reduce common cause failures of emergency power suppliers due to natural hazards. Thus, the CPG can provide a new mean for supplying emergency electric power during station blackout events of nuclear power plants. For this application, the CPG requirements are described with a typical configuration at the ocean side of a submerged tunnel.

• 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.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.319-326
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• 2018

Since the Fukushima Daiichi nuclear disaster, concern and worry about multiunit accidents have been increasing. Korea has a higher urgency to evaluate its site risk because its number of nuclear power plants (NPPs) and population density are higher than those in other countries. Since the 1980s, technical documents have been published on multiunit probabilistic safety assessment (PSA), but the Fukushima accident accelerated research on multiunit PSA. It is therefore necessary to summarize the present situation and draw implications for further research. This article reviews journal and conference papers on multiunit or site risk evaluation published between 2011 and 2016. The contents of the reviewed literature are classified as research status, initiators, and methodologies representing dependencies, and the insights and conclusions are consolidated. As of 2017, the regulatory authority and nuclear power utility have launched a full-scale project to assess multiunit risk in Korea. This article provides comprehensive reference materials on the necessary enabling technology for subsequent studies of multiunit or site risk assessment.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.738-745
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• 2019

Coprecipitation using hydrous ferric oxide (HFO) has been effectively used for the removal of radionuclides from radioactive wastewater. This work studied the dynamic behavior of HFO floc formation during the neutralization of acidic ferric iron in the presence of several radionuclides by using a photometric dispersion analyzer (PDA). Then the coagulation-flocculation system using HFO-anionic poly acrylamide (PAM) composite floc system was evaluated and compared in seawater and distilled water to find the effective condition to remove the target nuclides (Co-60, Mn-54, Sb-125, and Ru-106) present in wastewater generated in the severe accident of nuclear power plant like Fukushima Daiichi case. A ferric iron dosage of 10 ppm for the formation of HFO was suitable in terms of fast formation of HFO flocs without induction time, and maximum total removal yield of radioactivity from the wastewater. The settling time of HFO flocs was reduced by changing them to HFO-PAM composite floc. The optimal dosage of anionic PAM for HFO-anionic PAM floc system was approximately 1-10 ppm. The total removal yield of Mn-54, Co-60, Sb-125, Ru-106 radionuclides by the HFO-anionic PAM coagulation-flocculation system was higher in distilled water than in seawater and was more than 99%.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1742-1756
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• 2023

The hydrogen behavior in a nuclear containment vessel is a significant issue when discussing the potential of hydrogen combustion during a severe accident. After the Fukushima-Daiichi accident in Japan, we have investigated in-depth the hydrogen transport mechanisms by utilizing experimental and numerical approaches. Computational fluid dynamics is a powerful tool for better understanding the transport behavior of gas mixtures, including hydrogen. This paper describes a Large-eddy simulation of gas mixing driven by a high-buoyancy flow. We focused on the interaction behavior of heat and mass transfers driven by the horizontal high-buoyant flow during density stratification. For validation, the experimental data of the Containment InteGral effects Measurement Apparatus (CIGMA) facility were used. With a high-power heater for the gas-injection line in the CIGMA facility, a high-temperature flow of approximately 390 ℃ was injected into the test vessel. By using the CIGMA facility, we can extend the experimental data to the high-temperature region. The phenomenological discussion in this paper helps understand the heat and mass transfer induced by the high-buoyancy flow in the containment vessel during a severe accident.

• 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.