원자력발전소(원전) 시스템 내진성능 평가를 위하여 구조물내응답스펙트럼(ISRS)은 필수적으로 요구된다. 특히, 원전 부지 고유 스펙트럼 변경 시 새로운 ISRS 도출이 요구될 경우 지진 재해석 등의 상당한 비용을 필요로 하게 된다. 따라서 이 연구는 지진 재해석 이 필요 없는 ISRS 스케일링 근사 방법에 대한 여러 가지 접근법을 제공한다. 이러한 접근법으로 도출한 ISRS는 정확한 ISRS와 비교 한다. 근사 방법의 ISRS 가 원전 주요 시스템 지진응답 및 내진성능에 미치는 영향을 분석한다. 결과적으로 본 연구에서 제시한 ISRS 스케일링 근사 방법은 저주파에서 비교적 유사하게 ISRS를 도출하지만, 고주파에서는 그 정확도가 감소하였다. ISRS 스케일링 근사 방법이 시스템 지진응답/내진성능 산출 정확도에 미치는 영향은 방법의 시스템 주요 모드 응답 유사도 산출 정도에 따라 결정된 것을 확인할 수 있었다.

As climate change and population growth raise the likelihood of natural disasters, it becomes crucial to comprehend and mitigate these risks in vital infrastructure systems, especially nuclear power plants (NPPs). This research addresses the necessity for evaluating multiple hazards by concentrating on slope failures triggered by earthquakes near NPPs over a timeframe extending up to a return period of 100,000 years. Utilizing a Geographical Information System (GIS) and Monte Carlo Simulation (MCS), the research conducts a comprehensive fragility assessment to predict failure probability under varying ground-shaking conditions. According to the Newmark displacement method, factors such as Peak Ground Acceleration (PGA), slope angle, soil properties, and saturation ratio play significant roles in determining slope safety outcomes. The investigation aims to enhance understanding seismic event repercussions on NPP-adjacent landscapes, providing insights into long-term dynamics and associated risks. Results indicate an increase in slope vulnerability with longer return periods, with distinct instances of slope failures at specific return periods. This analysis not only highlights immediate seismic impacts but also underscores the escalating risk of slope displacement across the extended return period scales, crucial for evaluating long-term stability and associated hazards near nuclear infrastructure.

The importance of Structural Health Monitoring (SHM) in the industry is increasing due to various loads, such as earthquakes and wind, having a significant impact on the performance of structures and equipment. Estimating responses is crucial for the effective health management of these assets. However, using numerous sensors in facilities and equipment for response estimation causes economic challenges. Additionally, it could require a response from locations where sensors cannot be attached. Digital twin technology has garnered significant attention in the industry to address these challenges. This paper constructs a digital twin system utilizing the Long Short-Term Memory (LSTM) model to estimate responses in a pipe system under simultaneous seismic load and arbitrary loads. The performance of the data-driven digital twin system was verified through a comparative analysis of experimental data, demonstrating that the constructed digital twin system successfully estimated the responses.

고성능 콘크리트(HPC) 압축강도는 추가적인 시멘트질 재료의 사용으로 인해 예측하기 어렵고, 개선된 예측 모델의 개발이 필수적 이다. 따라서, 본 연구의 목적은 배깅과 스태킹을 결합한 앙상블 기법을 사용하여 HPC 압축강도 예측 모델을 개발하는 것이다. 이 논 문의 핵심적 기여는 기존 앙상블 기법인 배깅과 스태킹을 통합하여 새로운 앙상블 기법을 제시하고, 단일 기계학습 모델의 문제점을 해결하여 모델 예측 성능을 높이고자 한다. 단일 기계학습법으로 비선형 회귀분석, 서포트 벡터 머신, 인공신경망, 가우시안 프로세스 회귀를 사용하고, 앙상블 기법으로 배깅, 스태킹을 이용하였다. 결과적으로 본 연구에서 제안된 모델이 단일 기계학습 모델, 배깅 및 스태킹 모델보다 높은 정확도를 보였다. 이는 대표적인 4가지 성능 지표 비교를 통해 확인하였고, 제안된 방법의 유효성을 검증하였다.

In Korea, most nuclear power plants were designed based on the design response spectrum of Regulatory Guide 1.60 of the NRC. However, in the case of earthquakes occurring in the country, the characteristics of seismic motions in Korea and the design response spectrum differed. The seismic motion in Korea had a higher spectral acceleration in the high-frequency range compared to the design response spectrum. The seismic capacity may be reduced when evaluating the seismic performance of the equipment with high-frequency earthquakes compared with what is evaluated by the design response spectrum for the equipment with a high natural frequency. Therefore, EPRI proposed the inelastic energy absorption factor for the equipment anchorage. In this study, the seismic performance of welding anchorage was evaluated by considering domestic seismic characteristics and EPRI's inelastic energy absorption factor. In order to reflect the characteristics of domestic earthquakes, the uniform hazard response spectrum (UHRS) of Uljin was used. Moreover, the seismic performance of the equipment was evaluated with a design response spectrum of R.G.1.60 and a uniform hazard response spectrum (UHRS) as seismic inputs. As a result, it was confirmed that the seismic performance of the weld anchorage could be increased when the inelastic energy absorption factor is used. Also, a comparative analysis was performed on the seismic capacity of the anchorage of equipment by the welding and the extended bolt.

원자력발전소 기기 내진설계 및 지진해석은 비연계모델을 대상으로 수행된다. 그러나 이러한 비연계해석은 실제 구조물-기기 간 상호작용 등의 실제 현상을 모사할 수 없기 때문에 연계해석에 비하여 정확하지 못한 결과를 발생시키게 된다는 한계를 가진다. 이러 한 배경 아래 이 연구는 실제 원전 격납건물 구조물 및 관련 부계통을 대상으로 질량비와 고유진동수비를 고려하여 지진 연계해석과 비연계해석을 수행하고, 이를 바탕으로 부계통에서의 응답을 비교 분석하였다. 결과적으로 지진 연계해석 결과가 비연계해석 결과보 다 대다수 작은 값을 주는 것을 확인하였다. 이러한 결과는 기존 연구인 단순한 연계모델에 대한 해석 결과와 유사하지만, 부계통 응답 차이는 훨씬 더 두드러지게 나타나는 것을 확인하였다. 또한, 이는 지진파의 입력 주파수의 영향보다는 부계통의 설치위치에 영향을 받는 것으로 확인되었다. 마지막으로 비연계 및 연계 지진해석의 차이가 부계통의 질량비가 크고, 고유진동수가 거의 일치하는 영역 에서 발생하는 이유는 이 영역에서 주계통과 부계통 동적 상호작용이 크게 나타나기 때문인 것으로 보인다.

후쿠시마 원전사고 발생으로 다수기의 지진안전성에 관한 연구의 필요성이 부각되었다. 한 부지에 건설된 원자력발전소의 경우 유 사한 지진응답을 보이기 때문에 적게나마 원자력발전소 SSCs간의 지진손상에 대하여 상관성이 존재하므로 합리적 지진안전성 평가 를 위하여 지진손상 상관성을 고려하여야 한다. 본 연구에서는 쌍둥이 호기의 필수전원상실사건에 대하여 확률론적 지진안전성 평가 를 수행하였다. 적절한 지진손상 상관계수를 도출하기 위하여 확률론적 지진응답해석을 수행하여 적용하였다. External Event Mensuration System 프로그램을 활용하여 다수기의 필수전원상실사건의 고장수목을 구성하여 지진취약도 및 지진리스크를 분석하 였다. 또한 SSCs간의 지진손상 상관성을 완전독립 및 완전종속으로 고려하여 비교 분석을 수행하였다.

Nuclear power plants in Korea were designed and evaluated based on the NRC's Regulatory Guide 1.60, a design response spectrum for nuclear power plants. However, it can be seen that the seismic motion characteristics are different when analyzing the Gyeongju earthquake and the Pohang earthquake that has recently occurred in Korea. Compared to the design response spectrum, seismic motion characteristics in Korea have a larger spectral acceleration in the high-frequency region. Therefore, in the case of equipment with a high natural frequency installed in a nuclear power plant, seismic performance may be reduced by reflecting the characteristics of domestic seismic motions. The failure modes of the equipment are typically structural failure and functional failure, with an anchorage failure being a representative type of structural failure. In this study, comparative analyses were performed to decide whether to consider the inelastic behavior of the anchorage or not. As a result, it was confirmed that the seismic performance of the anchorages could be increased by considering the inelastic behavior of an anchorage.

원자력 발전소에 지진격리장치를 설치하여 내진성능을 향상시킬 수 있다. 그러나 지진격리장치의 적용으로 지반과 구조물 사이에서 큰 상대 변위가 발생하게 된다. 따라서 지진격리된 구조물과 일반 구조물을 연결하는 연결배관시스템의 경우 지진리스크가 증가할 수 있다. 따라서 이러한 배관시스템의 지진취약도를 분석할 필요가 있다. 본 연구에서는 지진취약도 분석 을 위해 지진격리된 APR1400 원자력발전소와 주증기관을 대상으로 지진취약도를 분석하였다. 주증기관은 지진격리된 nuclear island의 보조 건물과 터빈 건물을 연결하는 인터페이스 배관이다. 지진취약도 분석을 위한 파괴모드는 누출관통균열로 정의하였다. 누출은 실험결과와 수치해석을 통해 손상지수로 정량화하여 취약도 분석을 위한 파괴기준으로 사용하였다. 파괴기준의 변동에 의한 취약도 곡선의 변동성을 확인하기 위하여 손상지수의 최솟값, 최댓값, 평균값 및 중앙값을 파괴기준으로 하여 지진취약도 곡선을 작성하였다.

원자력시설 SPRA 방법으로서 기기 사이 부분 종속 관계를 정확하게 고려하기 위하여 샘플링기반접근법이 개발된 바 있다. 그러나 이는 샘플링 기반 방법이므로 정확한 지진 리스크 산정을 위하여 많은 수의 샘플을 추출해야 하는 단점이 있다. 이에 따라 본 연구에서 는 기존 방법을 개선하기 위한 효과적인 방법을 제안한다. 본 연구에서 제안한 방법의 주요한 특징은 다음과 같다. 기존 샘플링방법인 몬테카를로샘플링(MCS) 방법을 대신하여 다차원에서 효과적인 샘플링이 가능한 라틴하이퍼큐브샘플링(LHS) 방법을 샘플링기반 SPRA에 도입한다. 또한, 기존 지진세기 세분화 정도를 최종 지진 리스크 결과와 연계하여 결정한다. 제안된 방법이 결합된 샘플링기반 SPRA 접근법을 실제 원전 예제에 적용한 결과, 제안된 방법이 기존의 방법과 비교하여 결과 정확도에 있어서 거의 비슷하나 총 샘플 추출수 기준에서 효율성을 약 2배 가량 높이 것을 확인하였다. 또한, 샘플링 개수가 적은 영역에서 LHS 기반 방법이 MCS 기반 방법보다는 해의 정확도를 높이는 것을 확인할 수 있었다.

Nuclear power plant’s safety against seismic events is evaluated as risk values by probabilistic seismic safety assessment. The risk values vary by the seismic failure correlation between the structures, systems, and components (SSCs). However, most probabilistic seismic safety assessments idealized the seismic failure correlation between the SSCs as entirely dependent or independent. Such a consideration results in an inaccurate assessment result not reflecting real physical phenomenon. A nuclear power plant’s seismic risk should be calculated with the appropriate seismic failure correlation coefficient between the SSCs for a reasonable outcome. An accident scenario that has an enormous impact on a nuclear power plant’s seismic risk was selected. Moreover, the probabilistic seismic response analyses of a nuclear power plant were performed to derive appropriate seismic failure correlations between SSCs. Based on the analysis results, the seismic failure correlation coefficient between SSCs was derived, and the seismic fragility curve and core damage frequency of the loss of essential power event were calculated. Results were compared with the seismic fragility and core damage frequency of assuming the seismic failure correlations between SSCs were independent and entirely dependent.

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.

In this study, we develop a sampling-based seismic probabilistic risk assessment (SPRA) quantification technique that can accurately consider a partially dependent condition of component seismic fragility information. Specifically, the SPRA quantification method is proposed by combining the advantages of two representative methodologies: EPRI seismic fragility and JAERI seismic fragility input-based quantification. The most important feature of the proposed method is that it performs a SPRA using a sampling technique by transforming the EPRI seismic fragility input into JAERI seismic fragility input. When the proposed sampling-based approach was applied to an example of simple system and to a SPRA problem of a nuclear power plant, it was observed that the proposed method yields approximately similar system seismic fragility and seismic risk results as those of the exact solution. Therefore, it is believed that the approach proposed in this study can be used as a useful tool for accurately assessing seismic risks, considering the partial seismic dependence among the components; the existing SPRA method cannot handle such partial dependencies.

Seismic responses due to the dynamic coupling between a primary structure and secondary system connected to a structure are analyzed in this study. The seismic responses are compared based on dynamic coupling criteria and according to the error level in the natural frequency, with the recent criteria being reliant on the error level in the spectral displacement response. The acceleration responses and relative displacement responses of a primary structure and a secondary system for a coupled model and two different decoupled models of two degrees-of-freedom system are calculated by means of the time integration method. Errors in seismic responses of the uncoupled models are reduced with the recent criteria. As the natural frequency of the secondary system increases, error in the natural frequency decreases, but seismic responses of uncoupled models can be underestimated compared to that of coupled model. Results in this paper can help determine dynamic coupling and predict uncoupled models’ response conservatism.

최근 국내에서 발생한 경주지진 및 포항지진이 도심지 및 원자력발전소 주변에서 발생함에 따라 인근 주요 구조시스 템에 피해를 발생시킨 바 있다. 이에 따라, 본 연구는 원전 배관계통의 지진 거동 특성을 실험 및 해석적으로 규명하고, 이를 바탕으로 내진성능을 평가하는 연구를 수행하는 것을 목적으로 한다. 이를 위하여, 원전 배관계통을 대상으로 한 진동대 시험 결과를 바탕으로 배관 수치해석 모델을 수립하고 이를 시험 결과를 통하여 검증한다. 또한, 이러한 검증된 수치해석모델을 기반으로 배관계통의 내진성능을 평가한다. 수치해석 모델 검증 결과, 수립된 수치모델의 고유진동수, 감쇠비 및 변위 응답이 진동 대시험 결과와 유사한 것으로 나타났다. 이러한 검증된 수치모델을 바탕으로 평가된 내진성능 값은 기존 원전 배관의 내진성능 값의 범위와 비교적 유사한 값으로 평가되었다. 이는 설계기준지진을 크게 상회하는 값이지만, 원전 주요 기기 임을 감안할 때 추가적인 내진성능 상향이 확보되면 원전 지진 안전성 향상에 많은 기여를 할 수 있을 것으로 판단된다. 본 연구의 결과는 추후 원전 배관계통 내진성능평가를 위한 정량적인 자료로 활용이 가능할 것으로 판단된다.

In 2016, an earthquake occurred at Gyeongju, Korea. At the Wolsong site, the observed peak ground acceleration was lower than the operating basis earthquake (OBE) level of Wolsong nuclear power plant. However, the measured spectral acceleration value exceeded the spectral acceleration of the operating-basis earthquake (OBE) level in some sections of the response spectrum, resulting in a manual shutdown of the nuclear power plant. Analysis of the response spectra shape of the Gyeongju earthquake motion showed that the high-frequency components are stronger than the response spectra shape used in nuclear power plant design. Therefore, the seismic performance evaluation of structures and equipment of nuclear power plants should be made to reflect the characteristics of site-specific earthquakes. In general, the floor response spectrum shape at the installation site or the generalized response spectrum shape is used for the seismic performance evaluation of structures and equipment. In this study, a generalized response spectrum shape is proposed for seismic performance evaluation of structures and equipment for nuclear power plants. The proposed response spectrum shape reflects the characteristics of earthquake motion in Korea through earthquake hazard analysis, and it can be applied to structures and equipment at various locations.

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.

Severe earthquakes can cause damage to society both socially and economically. An appropriate initial response can alleviate damage from severe earthquakes. In order to formulate an appropriate initial response, it is necessary to identify damage situations in societies; however, it is difficult to grasp this information immediately after an earthquake event. In this study, an earthquake damage assessment methodology for buildings is proposed for estimating damage situations immediately after severe earthquakes. A response spectrum database is constructed to provide response spectra at arbitrary locations from earthquake measurements immediately after the event. The fragility curves are used to estimate the damage of the buildings. Earthquake damage assessment is performed from the response spectrum database at the building scale to provide enhanced damage condition information. Earthquake damage assessment for Gyeongju city and Pohang city were conducted using the proposed methodology, when an earthquake occurred on September 12, 2016, and November 15, 2017. Results confirm that the proposed earthquake damage assessment effectively represented the earthquake damage situation in the city to decide on an appropriate initial response by providing detailed information at the building scale.

The probabilistic seismic safety assessment is one of the methodology to evaluate the seismic safety of the nuclear power plants. The site characteristics of the nuclear power plant should be reflected when evaluating the seismic safety of the nuclear power plant. The Korea seismic characteristics are strong in high frequency region and may be different from NRC Regulatory Guide 1.60, which is the design spectrum of nuclear power plants. In this study, seismic response of a nuclear power plant structure by Pohang earthquake (2017.11.15. (KST)) is investigated. The Pohang earthquake measured at the Cheongsong seismic observation station (CHS) is scaled to the peak ground acceleration (PGA) of 0.2 g and the seismic acceleration time history curve corresponding to the design spectrum is created. A nuclear power plant of the containment building and the auxiliary buildings are modeled using OPENSEES to analyze the seismic response of the Pohang earthquake. The seismic behavior of the nuclear power plant due to the Pohang earthquake is investigated. And the seismic performances of the equipment of a nuclear power plant are evaluated by the HCLPF. As a result, the seismic safety evaluation of nuclear power plants should be evaluated based on site-specific characteristics of nuclear power plants.