• 한국전산구조공학회논문집
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• 제20권4호
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• pp.463-468
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• 2007

최근 일본에서 발생한 지진으로 이동통신용 전자장비들이 많이 파손되었으나. 국내에서 제작되어 일본에 설치된 모제품은 강진에서도 대부분 구조적 손상이 없었다. 본 논문에서는 이동통신 네트워크 장비의 정/동적 특성을 평가하기 위하여 제품의 유한요소 모델을 생성하여 정적 및 동적 구조해석을 수행하였다. 또한 Zone 3 GR-63-CORE 동적 실험을 수행하여 제품의 안정성을 검증하였고, 유한요소 구조해석 결과와 비교하여 실험과 해석 결과가 유사함이 입증되었다. 구조해석 결과인 동적 응답특성은 실험보다 다소 크게 나왔으며, 부재의 특성파악을 위해 구조물의 치수를 설계변수로 하여 진동특성에 대한 민감도 해석으로 고유진동수에 민감한 부재를 판별하였고, 치수변경으로 경량화 설계치를 산출하였다. 경량화된 디자인의 동적 응답변위가 원래 디자인보다 작게 나와 최적화 결과가 유용할 것으로 보인다.

• Nuclear Engineering and Technology
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• 제27권3호
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• pp.292-300
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• 1995

이 논문은 새로운 동적 인간신뢰도 분석방법을 제시하였고, 사고관리 방안의 수행시 인간오류확률의 계산에 이 방법을 적용하였다. 기존의 다른 방법과 비교하기 위하여 PSA의 HRA수행시 가장 많이 사용되는 THERP, HCR, 및 SLIM-MAUD 방법론들의 특징을 논의하였다. 정전사고시 공동범람시키는 방안을 예제로 사용하였다. 이 방법은 Requirement와 Achievement의 연관개념에 기초하고 있다. Achievement 변수의 불확정성은 MAAP 전산코드와 Latin Hypercube Sampling 방법을 이용하여 결정하였고 Requirement 변수값은 운전원과의 면담을 통하여 얻었다. 이렇게 얻어진 변수들의 분포를 가지고 여러가지 시간값의 평균과 분산에 대하여 인간오류 확률값을 계산하였다. 이 방법은 매우 유연하여 사고관리 전략수행과 관련한 행위를 포함한 어떤 종류의 운전원 행위에도 적용가능 함을 보여주었다.

• Nuclear Engineering and Technology
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• 제52권10호
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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.

• 한국안전학회지
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• 제27권6호
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• pp.192-204
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• 2012

For a severe accident of nuclear power plant, an approach to estimation of the radiological source term using a severe accident code(MELCOR) has been proposed. Although the MELCOR code has a capability to estimate the radiological source term, it has been hardly utilized for the radiological consequence analysis mainly due to a lack of understanding on the relevant function employed in MELCOR and severe accident phenomena. In order to estimate the severe accident source term to be linked with the radiological consequence analysis, this study proposes 4-step procedure: (1) selection of plant condition leading to a severe accident(i.e., accident sequence), (2) analysis of the relevant severe accident code, (3) investigation of the code analysis results and post-processing, and (4) generation of radiological source term information for the consequence analysis. The feasibility study of the present approach to an early containment failure sequence caused by a fast station blackout(SBO) of a reference plant (OPR-1000), showed that while the MELCOR code has an integrated capability for severe accident and source term analysis, it has a large degree of uncertainty in quantifying the radiological source term. Key insights obtained from the present study were: (1) key parameters employed in a typical code for the consequence analysis(i.e., MACCS) could be generated by MELCOR code; (2) the MELOCR code simulation for an assessment of the selected accident sequence has a large degree of uncertainty in determining the accident scenario and severe accident phenomena; and (3) the generation of source term information for the consequence analysis relies on an expert opinion in both areas of severe accident analysis and consequence analysis. Nevertheless, the MELCOR code had a great advantage in estimating the radiological source term such as reflection of the current state of art in the area of severe accident and radiological source term.

• 시스템엔지니어링학술지
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• 제16권2호
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• pp.97-109
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• 2020

On March 11, 2011, an earthquake followed by a tsunami caused an extended station blackout (SBO) at the Fukushima Dai-ichi NPP Units. The accident was initiated by a total loss of both onsite and offsite electrical power resulting in the loss of the ultimate heat sink for several days, and a consequent core melt in some units where proper mitigation strategies could not be implemented in a timely fashion. To enhance the plant's coping capability, the Diverse and Flexible Strategies (FLEX) were proposed to append the Emergency Operation Procedures (EOPs) by relying on portable equipment as an additional line of defense. To assess the success window of FLEX strategies, all sources of uncertainties need to be considered, using a physics-based model or system code. This necessitates conducting a large number of simulations to reflect all potential variations in initial, boundary, and design conditions as well as thermophysical properties, empirical models, and scenario uncertainties. Alternatively, data-driven models may provide a fast tool to predict the success window of FLEX strategies given the underlying uncertainties. This paper explores the applicability of Artificial Intelligence (AI) to identify the success window of FLEX strategy for extended SBO. The developed model can be trained and validated using data produced by the lumped parameter thermal-hydraulic code, MARS-KS, as best estimate system code loosely coupled with Dakota for uncertainty quantification. A Systems Engineering (SE) approach is used to plan and manage the process of using AI to predict the success window of FLEX strategies under extended SBO conditions.

• 시스템엔지니어링학술지
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• 제16권2호
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• pp.27-37
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• 2020

The Fukushima Daiichi accident in 2011 revealed some vulnerabilities of existing Nuclear Power Plants (NPPs) under extended Station Blackout (SBO) accident conditions. One of the key Severe Accident Management (SAM) strategies developed post Fukushima accident is the In-Vessel Retention (IVR) Strategy which aims to retain the structural integrity of the Reactor Pressure Vessel (RPV). RELAP/SCDAPSIM/MOD3.4 is selected to predict the thermal-hydraulic response of APR1400 undergoing an extended SBO. To assess the effectiveness of the IVR strategy, it is essential to quantify the underlying uncertainties. In this work, both the epistemic and aleatory uncertainties are considered to identify the success window of the IVR strategy. A set of in-vessel relevant phenomena were identified based on Phenomena Identification and Ranking Tables (PIRT) developed for severe accidents and propagated through the thermal-hydraulic model using Wilk's sampling method. For this work, a Systems Engineering (SE) approach is applied to facilitate the development process of assessing the reliability and robustness of the APR1400 IVR strategy. Specifically, the Kossiakoff SE method is used to identify the requirements, functions and physical architecture, and to develop a design verification and validation plan. Using the SE approach provides a systematic tool to successfully achieve the research goal by linking each requirement to a verification or validation test with predefined success criteria at each stage of the model development. The developed model identified the conditions necessary for successful implementation of the IVR strategy which maintains the vessel integrity and prevents a melt-through.

• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• 대한인간공학회지
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• 제32권1호
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• pp.117-124
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• 2013

Objective: This paper presents additional considerations related to organization and safety culture extracted from recent human error incidents in Korea, such as station blackout(i.e., SBO) in Kori#1. Background: Safety culture has been already highlighted as a major cause of human errors after 1986 Chernobyl accident. After Fukushima accident in Japan, the public acceptance for nuclear energy has taken its toll. Organizational characteristics and culture became elucidated as a major contributor again. Therefore many nuclear countries are re-evaluating their safety culture, and discussing any preparedness and its improvement. On top of that, there was an SBO in 2012 in the Kori#1. Korean public feels frustrated due to the similar human errors causing to a catastrophe like Fukushima accident. Method: This paper reassesses Japan's incidents, and revisits Korea's recent incidents. It focuses on the analysis of the hazards rather than the causes of human errors, the derivation of countermeasures, and their implementation. The preceding incidents and conclusions from Japanese experience are also re-analyzed. The Fukushima accident was an SBO due to the natural disaster such as earthquakes and a successive tsunami. Unlike the Fukushima accident, the Kori#1 incident itself was simple and restored without any loss and radioactive release. However, the fact that the incident was deliberately concealed led to massive distrust. Moreover, the continued violation of rules and organized concealment of the accident are serious signs of a new distorted type of human errors, blatantly revealing the cultural and fundamental weakness of the current organization. Result: We should learn from Japanese experiences who had taken pride in its safety technology and fairly high confidence in safety culture. Japan's first criticality accident in JCO facility splashed cold water on that confidence. It has turned out to be a typical case revealing the problems in the organization and safety culture. Since Japan has failed to gain lessons and countermeasure, the issue persists to the Fukushima incident. Conclusion: Safety culture is not a specific independent element, which makes it difficult to either evaluate it properly or establish countermeasures from the lessons. It may continue to expose similar human errors such as concealment of incident and manipulation of bad data. Application: Not only will this work establish the course of research for organization and safety culture, but this work will also contribute to the revitalization of Korea's nuclear industry from the disappointment after the export contract to UAE.

• 한국인터넷방송통신학회논문지
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• 제17권2호
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• pp.245-250
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• 2017

케이블은 설치하여 동작함과 동시에 열화과정이 진행된다. 고압전력케이블 시스템에 있어, 케이블의 절연층 및 방식층의 절연상태뿐만 아니라 케이블을 연결하는 단말부 및 접속부의 절연상태도 함께 감시하여야 케이블 시스템의 절연상태를 종합 관리할 수 있다. 케이블 시스템(케이블 자체 및 접속재)의 상태가 계속 나빠지는 경우, XLPE의 절연 파괴현상으로 인한 화재가 발생한다. 우리는 케이블시스템의 절연 상태를 감시하기 위한 장비를 개발하여 충남 태안의 한국 서부발전주식회사(Korea Western Power Co. Ltd.)에 설치하였다. 이 논문에서, 이 장비를 사용하여 케이블에 흐르는 누설전류를 추출하여 누설전류의 표준편차를 계산하여 분석한 결과를 제시한다. 정해진 누설 전류의 표준 편차값이 기준값 미만이면 안전하지만. 그 이상이면 케이블시스템의 단말부 및 접속부의 절연상태가 나쁜 것으로 판단하고, 새로운 케이블시스템의 단말부 및 접속부로 대체함으로써 전력공급 중단으로 인한 전력설비의 중단사고를 미연에 방지할 수 있다.