설계초과 지진에 대한 원전 지진안전성 평가기술 고찰 및 제언

최인길,Choi, In-Kil 한국압력기기공학회 2017 한국압력기기공학회 논문집 Vol.13 No.1

After Kyeongju earthquake occurred in September 12, 2016, the seismic safety of nuclear power plants became important issue in our country. The seismic safety of nuclear power plant against beyond design basis earthquake became very important to secure the public safety. In this paper, the current status of the seismic safety assessment methodology is reviewed and some aspects for the reliability improvement of the seismic safety assessment results are proposed. Seismic margin analysis and probabilistic seismic safety assessment have been used for the seismic safety evaluation of a nuclear power pant. The basic procedure and the related issues and proposals for the probabilistic seismic safety assessment are investigated.

다중계통모델의 지진리스크 산정용 프로그램 개발

최인길(Choi In-Kil),김도연(Kim Do-Yeon),박진희(Park Jin-Hee) 대한토목학회 2009 대한토목학회 학술대회 Vol.2009 No.10

후쿠오카 지진의 지진동 감쇄 특성 평가

최인길(Choi In-Kil),中島正人(Masato Nakajima),전영선(Choun Young-Sun),大鳥靖樹(Yasuki Ohtori),연관희(Yun Kwan-Hee) 대한토목학회 2006 대한토목학회 학술대회 Vol.2006 No.10

구조물 및 기기의 내진성능 평가를 위한 고주파수 지진에 의한 원자력발전소의 지진응답 증폭계수

임승현,최인길,전법규,곽신영,Eem, Seung-Hyun,Choi, In-Kil,Jeon, Bub-Gyu,Kwag, Shinyoung 한국지진공학회 2020 한국지진공학회논문집 Vol.24 No.3

Analysis of the 2016 Gyeongju earthquake and the 2017 Pohang earthquake showed the characteristics of a typical high-frequency earthquake with many high-frequency components, short time strong motion duration, and large peak ground acceleration relative to the magnitude of the earthquake. Domestic nuclear power plants were designed and evaluated based on NRC's Regulatory Guide 1.60 design response spectrum, which had a great deal of energy in the low-frequency range. Therefore, nuclear power plants should carry out seismic verification and seismic performance evaluation of systems, structures, and components by reflecting the domestic characteristics of earthquakes. In this study, high-frequency amplification factors that can be used for seismic verification and seismic performance evaluation of nuclear power plant systems, structures, and equipment were analyzed. In order to analyze the high-frequency amplification factor, five sets of seismic time history were generated, which were matched with the uniform hazard response spectrum to reflect the characteristics of domestic earthquake motion. The nuclear power plant was subjected to seismic analysis for the construction of the Korean standard nuclear power plant, OPR1000, which is a reactor building, an auxiliary building assembly, a component cooling water heat exchanger building, and an essential service water building. Based on the results of the seismic analysis, a high-frequency amplification factor was derived upon the calculation of the floor response spectrum of the important locations of nuclear power plants. The high-frequency amplification factor can be effectively used for the seismic verification and seismic performance evaluation of electric equipment which are sensitive to high-frequency earthquakes.

기기의 지진취약도 평가를 위한 구조물 비탄성구조응답계수의 재평가

박준희,최인길,Park, Junhee,Choi, In-Kil 한국전산구조공학회 2015 한국전산구조공학회논문집 Vol.28 No.3

There are a lot of equipment related to safety and electric power production in nuclear power plants. The structure and equipment in NPPs were generally designed considering a high safety factor to remain in the elastic zone under earthquake load. However it is needed to revaluate the seismic capacity of the structure and equipment as the magnitude of earthquake was recently increased. In this study the floor response due to the nonlinear behaviors of structure was analyzed and the inelastic structural response factor was calculated by the nonlinear time history analysis. The inelastic structural response factor was calculated by the EPRI method and the nonlinear analysis method to realistically evaluate the seismic fragility for the equipment. According to the analysis result, it was represented that the inelastic structural response factor was affected by the natural frequency of equipment, the location of equipment and the dynamic property of structure. 원자력발전소에는 전력생산과 안전과 관련된 수많은 기기들이 존재하고 있다. 기본적으로 원자력발전소의 구조물과 기기는 지진시 탄성거동올 목표로 안전율을 매우 높게 적용하여 설계해 왔다. 그러나 최근 발생한 지진의 규모가 증가함에 따라 설계수준을 초과한 지진에 대한 기기의 안전성을 재평가할 필요가 있다. 본 연구에서는 구조물의 비선형 거동에 의한 층응답을 분석하였고, 비선형해석에 의한 구조물의 비탄성구조응답계수를 재평가하였다. 기기의 지진취약도 평가시 구조물의 비탄성구조응답이 어떤 영향을 주는지 분석하기 위하여 재평가된 구조물의 비탄성구조응답계수와 기존에 사용되어온 구조물 비탄성구조응답계수를 적용하여 지진취약도 평가를 수행하였다. 해석결과에 따르면 비탄성구조응답계수는 기기의 고유진동수, 기기의 위치 그리고 구조물의 동특성에 따라 영향을 받는 것으로 나타났다.

원전 전기캐비넷의 지진취약도 재평가를 위한 진동대 실험

김민규,최인길,Kim, Min-Kyu,Choi, In-Kil 한국전산구조공학회 2011 한국전산구조공학회논문집 Vol.24 No.3

본 연구에서는 원자력발전소의 주요 설비중의 하나인 전기설비를 대상으로 지진취약도 재평가를 위한 진동대 실험을 수행하였다. 원자력발전소 내에는 많은 전기설비들이 설치되어 있으며, 이러한 전기설비의 손상은 전기설비 자체의 손상에서 그치는 것이 아니고 발전소 전체의 안전성에 큰 영향을 미칠 수 있다. 따라서 원자력발전소의 확률론적 지진안전성 평가에서는 주요 전기설비에 대한 지진취약도 결과를 활용한 평가를 수행하고 있다. 본 연구에서는 기존의 확률론적 지진안전성 평가에서 사용하고 있는 전기설비의 지진취약도 값에 대한 재평가를 위하여 원자력발전소에서 사용하고 있는 주요 기기에 대한 진동대 실험을 수행하였다. 평가대상 전기설비로는 480V MCC를 선정하였으며, 진동대 실험을 위하여 NRC 설계지진, 등재해도 스펙트럼에 의한 인공지진 그리고 PAB165'에서의 층응답스펙트럼을 이용한 인공지진의 3가지 지진파를 이용하였다. 설계지진동 수준인 최대지반가속도 0.2g부터 단계적으로 입력수준을 증가시키면서 실험을 수행하였다. NUREG/CR-5203에서 제시하고 있는 방법에 의거하여 캐비넷에서의 증폭비를 비교하였으며, EPRI TR-103959의 방법으로 취약도 평가를 수행하여 기존의 확률론적 지진안전성 평가에서 사용하고 있는 지진취약도 결과와 비교하였다. 결론적으로 기존의 보고서에서 제시하고 있는 취약도 결과가 다소 보수적으로 평가하고 있음을 알 수 있었다. In this study, a seismic behavior of electrical cabinet system in Nuclear Power Plants(NPPs) was evaluated by the shaking table test. A 480V Motor Control Centers(MCCs) was selected for the shaking table test, and a real MCC cabinet for the Korea Nuclear Power Plant site was rented by manufactured company. For the shaking table tests, three kinds of seismic input motions were used, which were a US NRC Reg. guide 1.60 design spectrum, a UHS spectrum and PAB 165' floor response spectrum(FRS). Especially, the UHS input motion was selected for an evaluation of structural seismic amplification effects, three directional accelerations were measured at three points outside on the cabinet system and also that of the incabinet response amplification, accelerations were measure at two points which were mounted in electrical equipment such as relay. Seismic amplification effect is determined at the outside and inside of a cabinet as input seismic motion, and compared to the results which are calculated by analytical method based on NUREG/CR-5203.

비선형 유한요소해석을 이용한 CANDU형 격납건물의 내압취약도 평가

함대기,최인길,이홍표,Hahm, Dae-Gi,Choi, In-Kil,Lee, Hong-Pyo 한국전산구조공학회 2010 전산구조공학 Vol.23 No.4

CANDU형 격납건물에 대하여 극한내압하중에 대한 확률론적 취약도 평가를 수행하였다. 격납건물 성능의 불확실성은 가동중 검사 결과를 통해 얻어진 재료 물성치 중앙값과 텐던 긴장력 중앙값을 적용하여 고려하였다. 격납건물은 개구부를 고려하여 3차원 유한요소로 모델링하였으며, 확률론적 취약도 평가를 위하여 대규모의 비선형 유한요소해석 모델을 적용하기에 적합한 효율적인 취약도 평가기법을 개발하였다. 월성 1호기 격납건물에 대한 물성치를 사용하였다. 개발된 새로운 취약도 평가기법을 도입하여 각각의 파괴모드에 대한 취약도 평가를 수행하였으며, 파괴모드 별, 신뢰도 수준별 취약도 곡선을 도출하였다. 벽체 중단부가 극한내압발생으로 인한 방사능물질 누출에 가장 취약한 것으로 나타났다. In this paper an assessment of the internal pressure fragility of the CANDU type containment buildings is performed. The uncertainties of the performance of the containment buildings, material properties and tendon characteristics are referred from the in-service reports of Wolsung Unit 1. The containment buildings are modeled as a three-dimensional finite elements with considering the major opening and penetrations. A new method to evaluate the probabilistic fragility of the massive structural system is developed. The fragility curves of the target containment building are presented with repect to the failure modes and reliability levels. The center of wall is reveled as the most weak structural component of the containment building in the sense of the rupture and catastrophic rupture failure modes.

비선형 유한요소해석을 이용한 CANDU형 격납건물의 내압취약도 평가

함대기,최인길,이홍표,Hahm, Dae-Gi,Choi, In-Kil,Lee, Hong-Pyo 한국전산구조공학회 2010 한국전산구조공학회논문집 Vol.23 No.4

CANDU형 격납건물에 대하여 극한내압하중에 대한 확률론적 취약도 평가를 수행하였다. 격납건물 성능의 불확실성은 가동중 검사 결과를 통해 얻어진 재료 물성치 중앙값과 텐던 긴장력 중앙값을 적용하여 고려하였다. 격납건물은 개구부를 고려하여 3차원 유한요소로 모델링하였으며, 확률론적 취약도 평가를 위하여 대규모의 비선형 유한요소해석 모델을 적용하기에 적합한 효율적인 취약도 평가기법을 개발하였다. 월성 1호기 격납건물에 대한 물성치를 사용하였다. 개발된 새로운 취약도 평가기법을 도입하여 각각의 파괴모드에 대한 취약도 평가를 수행하였으며, 파괴모드 별, 신뢰도 수준별 취약도 곡선을 도출하였다. 벽체 중단부가 극한내압발생으로 인한 방사능물질 누출에 가장 취약한 것으로 나타났다. In this paper an assessment of the internal pressure fragility of the CANDU type containment buildings is performed. The uncertainties of the performance of the containment buildings, material properties and tendon characteristics are referred from the in-service reports of Wolsung Unit 1. The containment buildings are modeled as a three-dimensional finite elements with considering the major opening and penetrations. A new method to evaluate the probabilistic fragility of the massive structural system is developed. The fragility curves of the target containment building are presented with repect to the failure modes and reliability levels. The center of wall is reveled as the most weak structural component of the containment building in the sense of the rupture and catastrophic rupture failure modes.

다중기기 손상 상관성에 의한 지진리스크 영향 분석

임승현,최인길,Eem, Seung-Hyun,Choi, In-Kil 한국지진공학회 2019 한국지진공학회논문집 Vol.23 No.2

The seismic safety of nuclear power plants has always been emphasized by the effects of accidents. In general, the seismic safety evaluation of nuclear power plants carries out a seismic probabilistic safety assessment. The current probabilistic safety assessment assumes that damage to the structure, system, and components (SSCs) occurs independently to each other or perfect dependently to each other. In case of earthquake events, the failure event occurs with the correlation due to the correlation between the seismic response of the SSCs and the seismic performance of the SSCs. In this study, the EEMS (External Event Mensuration System) code is developed which can perform the seismic probabilistic safety assessment considering correlation. The developed code is verified by comparing with the multiplier n, which is for calculating the joint probability of failure, which is proposed by Mankamo. It is analyzed the changes in seismic fragility curves and seismic risks with correlation. As a result, it was confirmed that the seismic fragility curves and seismic risk change according to the failure correlation coefficient. This means that it is important to select an appropriate failure correlation coefficient in order to perform a seismic probabilistic safety assessment. And also, it was confirmed that carrying out the seismic probabilistic safety assessment in consideration of the seismic correlation provides more realistic results, rather than providing conservative or non-conservative results comparing with that damage to the SSCs occurs independently.

경년열화된 철근콘크리트 전단벽의 지진응답에 영향을 미치는 변수들의 민감도분석

박준희,전영선,최인길,Park, Jun-Hee,Choun, Young-Sun,Choi, In-Kil 한국전산구조공학회 2011 한국전산구조공학회논문집 Vol.24 No.4

철근콘크리트 구조물은 타설 후 시간이 경과함에 따라 물리적인 요인과 화학적인 요인으로 인하여 열화가 진행된다. 열화된 구조물의 효율적인 구조해석을 수행하기 위하여 구조물의 거동과 밀접한 관련이 있는 중요열화변수를 정의하는 것은 필요하다. 본 연구에서는 경년열화된 철근콘크리트 전단벽의 중요변수를 분석하기 위하여 일계이차모멘트법을 이용하여 민감도해석을 수행하였다. 콘크리트의 경화현상을 고려하지 않을 경우 구조물의 열화성능이 과소평가될 수 있으므로 콘크리트의 경화에 따른 해석변수들의 민감도를 분석하였다. 열화된 전단벽에서 변수의 중요도는 토네이도 다이어그램으로 나타내었다. After a concrete is poured, reinforced concrete structures were distressed by physical and chemical factor over time. It is in need to define important variables related to structural behavior for effectively conducting seismic analysis of structures with age-related degradation. In this study, a sensibility analysis using the first-order second moment method was performed to analyze an important variables for the reinforced concrete shear wall with age-related degradation. Because the seismic capacity of aging structures without a concrete hardening effect can be underestimated, the sensibility of analysis variables was analyzed according to the concrete hardening. Important variables for RC shear wall with age-related degradation was presented by using the tornado diagram.