• Nuclear Engineering and Technology
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• 제53권3호
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• pp.967-973
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• 2021

In 2011, an earthquake and subsequent tsunami hit the Fukushima Daiichi Nuclear Power Plant, causing simultaneous accidents in several reactors. This accident shows us that if there are several reactors on site, the seismic risk to multiple units is important to consider, in addition to that to single units in isolation. When a seismic event occurs, a seismic-failure correlation exists between the nuclear power plant's structures, systems, and components (SSCs) due to their seismic-response and seismic-capacity correlations. Therefore, it is necessary to evaluate the multi-unit seismic risk by considering the SSCs' seismic-failure-correlation effect. In this study, a methodology is proposed to obtain the seismic-response-correlation coefficient between SSCs to calculate the risk to multi-unit facilities. This coefficient is calculated from a probabilistic multi-unit seismic-response analysis. The seismic-response and seismic-failure-correlation coefficients of the emergency diesel generators installed within the units are successfully derived via the proposed method. In addition, the distribution of the seismic-response-correlation coefficient was observed as a function of the distance between SSCs of various dynamic characteristics. It is demonstrated that the proposed methodology can reasonably derive the seismic-response-correlation coefficient between SSCs, which is the input data for multi-unit seismic probabilistic safety assessment.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.31-34
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• 2013

According to the increase in numbers and operation time of domestic Nuclear Power Plants, KHNP(Korea Hydro & Nuclear Power) has many operating experiences. These show that most of the accidents repeatedly occurred not by the new sources or mechanism like the Fukushima Accident, but by the human and equipment errors from normal habits, process, design, maintenance etc.. These lessons show that the well-established systematic approach is requested to take lessons from past experiences. For this reason, developed countries established INPO, WANO, COG as a non-profit professional organizations to actively share their operating experiences. KHNP is also trying to promote the utilization of operating experiences. As part of this effort, KHNP is developing the KHNP-JIT, reflecting the overseas JIT and the domestic experiences.

• 지질공학
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• 제24권4호
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• pp.609-613
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• 2014

The Korean Peninsula, located at the southeastern tip of the Eurasian Plate, is known to be tectonically stable, and no critical evidence has yet been found that would override the safety design of nuclear facilities in South Korea. Because a nuclear power plant, like other major social overhead capital facilities, could cause great damage to both the environment and society through an unexpected tectonic event, even one of extremely low probability, like the Fukushima accident, a defense-in-depth safety approach is required in geological and geotechnical site safety evaluation for nuclear projects. This paper introduces the regulatory procedures that are in place to confirm nuclear site safety and site monitoring (e.g., earthquakes and groundwater) systems applied to nuclear facilities in order to reduce inherent uncertainties within the site safety review of geological and seismological issues related with a NPP project.

• 대한인간공학회지
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• 제35권2호
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• pp.75-84
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• 2016

Objective: The aim of this study is to suggest some improvement ideas based on the validity and the reliability analyses of the current safety culture measurement method applied to the Korean nuclear power industry. Background: Wrong safety culture is known as one of the major causes of the disasters such as the space shuttle Columbia disaster or the Fukushima Nuclear Power Plant accident. Assessment of safety culture of an organization is important to build a safer organizational environment as well as to identify the risks hidden in the organization. Method: A face validity of the current safety culture measurement method was analyzed by comparison of the key factors of safety culture in the Korean nuclear power industry with those factors reviewed in the previous studies. The current interview method was analyzed to identify the problems which degrade the consistency of evaluation. Results: Most safety culture factors reviewed in the literatures are covered in the list of the Korean nuclear power industry safety culture factors. However the unstructured questions used in the interview may result in inconsistency of safety culture evaluation among interviewers. Conclusion: This study suggests some examples which might improve the consistency of interviewers' evaluation on safety culture such as a post interview evaluation form. Application: An extended post interview evaluation form might help to increase the accuracy of the interviewing method for Korean nuclear industry safety culture evaluation.

• 전자공학회논문지
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• 제51권5호
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• pp.58-69
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• 2014

후쿠시마 사고와 같은 중대사고에 대비하기 위하여 극한환경 내에서도 동작하여 원자력 발전소 내부 상태에 관련된 계측신호를 취득하고, 사고복구에 필요한 밸브, 펌프 등과 같은 비상기기를 작동시킬 수 있는 극한환경용 제어기기와 발전소로부터 최소 30km 떨어진 곳에서 발전소 내부 상황을 감시하고, 제어할 수 있는 모바일 원격 제어실 등으로 구성된 비상대응시스템이 필요하다. 본 논문에서는 극한환경용 제어기기와 모바일 원격 제어실과의 연계를 위한 비상대응시스템 개념 설계 및 성능분석에 대하여 논의하였다. 비상통신시스템은 IEEE 802.11 기술표준을 이용한 지상망과 천리안 위성을 이용한 위성망의 이중화 시스템으로 구성되고, 각 시스템에 대하여 통신링크 버짓, Throughput, 지연시간을 분석하였다.

• 한국전산구조공학회논문집
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• 제26권6호
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• pp.415-421
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• 2013

본 논문에서는 면진시스템이 원전에 적용될 경우 원전구조물의 생애주기 성능에 미치는 영향을 소개한다. 최근 내진설계와 더불어 강진발생 예상 지역에 적용을 목적으로 개발되는 면진시스템은 구조물을 장주기화하여 응답가속도를 줄이고 상대변위를 늘려줌으로써 구조물의 안전성을 증진시키는 것으로 알려져 있다. 따라서, 구조물의 안전성이 중요시되는 원전구조물에 면진시스템을 적용하기 위한 연구가 국내에서 진행 중에 있다. 본 연구에서는 원전구조물의 생애주기 성능분석에 있어서 특징을 분석하고, 면진시스템이 적용될 경우 원전구조물의 생애주기성능에 있어서 미치는 영향을 평가함으로써, 도출된 결과를 면진시스템 적용의 정량적인 타당성 평가에 활용할 수 있다.

• 해양환경안전학회지
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• 제27권5호
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• pp.618-629
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• 2021

2011년 동일본대지진에 의해 발생한 후쿠시마 원자력 발전소 사고와 최근 국내 지진 발생 빈도의 증가는 원자력 발전소의 지진 안전성에 대한 불안감을 야기하였다. 더불어 최근(2021년) 일본 동경전력은 후쿠시마 원전 오염수의 태평양 방류를 결정하였으며, 이로 인해 국내외 수산물을 통한 방사능 오염 가능성이 높아지면서 국민들의 우려가 급증하고 있다. 후쿠시마 원전사고 이후 해양으로의 인공방사능 유입에 관한 연구는 국제적으로 많이 이루어졌으나, 한국인의 주요 식재료인 동아시아 연근해의 수산물에서 인공방사능의 분포 현황 및 축적에 대한 연구는 상대적으로 부족한 실정이다. 따라서 이 논문에서는 후쿠시마 원전사고 이후, 국내산 수산물에서의 원전 기원 인공방사능(예, ¹³⁷Cs, ^239+240Pu, ⁹⁰Sr 등)의 분포 특성과 관련한 최근 연구 사례들을 소개하고자 한다. 또한, 후쿠시마 원전오염수의 방류와 더불어 2030년까지 계획된 중국의 신규 원전 시설로 인한 향후 한반도 주변해역의 방사능 유출 영향에 대한 대비 및 사전 연구가 필요한 시점이기에 향후 연구 방향들을 제안하고자 한다.

• 한국안전학회지
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• 제35권5호
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• pp.121-127
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• 2020

Since the Fukushima nuclear accident in 2011, the public were concerned about the safety of Nuclear Power Plants (NPPs) in extreme natural disaster situations, such as earthquakes, flooding, heavy rain and tsunami, have been increasing around the world. Accordingly, the Stress Test was conducted in Europe, Japan, Russia, and other countries by reassessing the safety and response capabilities of NPPs in extreme natural disaster situations that exceed the design basis. The extreme natural disaster can put the NPPs in beyond-design-basis conditions such as the loss of the power system and the ultimate heat sink. The behaviors and capabilities of NPPs with losing their essential safety functions should be measured to find and supplement weak areas in hardware, procedures and coping strategies. The Loss of Ultimate Heat Sink (LUHS) accident assumes impairment of the essential service water system accompanying the failure of the component cooling water system. In such conditions, residual heat removal and cooling of safety-relevant components are not possible for a long period of time. It is therefore very important to establish coping strategies considering all available equipment to mitigate the consequence of the LUHS accident and keep the NPPs safe. In this study, thermal hydraulic behavior of the LUHS event was analyzed using RELAP5/Mod3.3 code. We also performed the sensitivity analysis to identify the effects of the operator recovery actions and operation strategy for charging pumps on the results of the LUHS accident.

• 한국안전학회지
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• 제37권5호
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• pp.80-88
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• 2022

The Korean nuclear industry had developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code, which adopts a two-fluid, three-field model that is comprised of gas, continuous liquid and droplet fields and has the capability to simulate three-dimensional models. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for the accident management plan of a nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification is required for the separate and integral effect experiments. Therefore, the goal of this work is to verify the calculation capability of the SPACE code for multiple failure accidents. For this purpose, an experiment was conducted to simulate a Control Element Drive Mechanism (CEDM) break with a safety injection failure using the ATLAS test facility, which is operated by Korea Atomic Energy Research Institute (KAERI). This experiment focused on the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The results of the overall system transient response using the SPACE code showed similar trends with the experimental results for parameters such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it can be concluded that the SPACE code has sufficient capability to simulate a CEDM break with a safety injection failure accident.