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.

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.

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.

For the important safety system, two or more units of identical equipment or redundant components with similar function were installed to prevent abnormal failure. If the failure probability of such equipment is independent, this redundancy could increase the system safety remarkably. However, if the failure of each component is highly correlated by installing in a structure or experiencing an earthquake event, the expected redundancy effect will decrease. Therefore, the seismic correlation of the equipment should be evaluated quantitatively for the seismic probabilistic safety assessment. The correlation effect can be explained in the procedure of constructing fragility curves. In this study, several methodologies to quantify the seismic correlation in the failure probability calculation for multiple components were reviewed and two possible ways considering the realistic situation were selected. Simple examples were tested to check the applicability of these methods. The conversion method between these two methods was suggested to render the evaluation using the advantages of each method possible.

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.

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.

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.

The tsunami hazard analysis is performed for testing the application of probabilistic tsunami hazard analysis to nuclear power plant sites in the Korean Peninsula. Tsunami hazard analysis is based on the seismic hazard analysis. Probabilistic method is adopted for considering the uncertainties caused by insufficient information of tsunamigenic fault sources. Logic tree approach is used. Uljin nuclear power plant (NPP) site is selected for this study. The tsunamigenic fault sources in the western part of Japan (East Sea) are used for this study because those are well known fault sources in the East Sea and had several records of tsunami hazards. We have performed numerical simulations of tsunami propagation for those fault sources in the previous study. Therefore we use the wave parameters obtained from the previous study. We follow the method of probabilistic tsunami hazard analysis (PTHA) suggested by the atomic energy society of Japan (AESJ). Annual exceedance probabilities for wave height level are calculated for the site by using the information about the recurrence interval, the magnitude range, the wave parameters, the truncation of lognormal distribution of wave height, and the deviation based on the difference between simulation and record. Effects of each parameters on tsunami hazard are tested by the sensitivity analysis, which shows that the recurrence interval and the deviation dominantly affects the annual exceedance probability and the wave heigh level, respectively.

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.

There has been an increasing demand for introducing a base isolation system to secure the seismic safety of a nuclear power plant. However, the design criteria and the safety assessment methodology of a base isolated nuclear facility are still being developed. A performance based design concept for the base isolation system needs to be added to the general seismic design procedures. For the base isolation system, the displacement responses of isolators excited by the extended design basis earthquake are important as well as the design displacement. The possible displacement response by the extended design basis earthquake should be limited less than the failure displacement of the isolator. The failure of isolators were investigated by an experimental test to define the ultimate strain level of rubber bearings. The uncertainty analysis, considering the variations of the mechanical properties of isolators and input ground motions, was performed to estimate the probabilistic distribution of the isolator displacement. The relationship of the displacement response by each ground motion level was compared in view of a period elongation and a reduction of damping. Finally, several examples of isolator parameters are calculated and the considerations for an acceptable isolation design is discussed.

원전의 항공기 충돌 리스크 평가에 사용되는 대표매개변수를 선정하기 위한 방법론을 개발하였다. 대상 원전은 국내의 대표적인 경수로형 원전 중 하나로 선정하여 3차원 유한요소 해석 모델을 구축하였다. 콘크리트 재료모델에는 소성손상모델이 적용되었으며, 강재는 다중선형곡선거동을 가지는 것으로 모델링하였다. 운동에너지, 전체 충격량, 최대 충격량, 최대 하중 등 4종의 대표매개변수 후보군을 선정하였다. 각각의 매개변수 후보군은 모두 충돌 속도와 질량의 함수로 표현되므로, 충돌속도 50~200m/s, 항공유량 30~90%의 범위에 대하여 매개변수값을 도출하고 충돌 해석을 수행하여, 충돌 시의 구조 응답과의 상관관계를 분석하였다. 모든 해석에서 항공기의 기종은 보잉767 기종으로 선정하였다. 충돌해석에는 Riera의 하중-시간이력 함수를 이용한 해석기법을 적용하였다. 매개변수와 충돌 시 응답의 상관관계 적합성은 결정계수값을 이용하여 분석하였다. 4 종의 대표매개변수 후보군 중 최대 하중값이 가장 직관적일 뿐만 아니라 본 연구에서의 해석 케이스에서는 응답과의 상관성도 가장 뛰어난 것으로 나타남에 따라, 항공기충돌 리스크 평가를 위하여 가장 적합한 매개변수라 할 수 있을 것으로 판단되었다.

면진장치는 지진력을 감소시키기 위하여 사용되어왔다. 원자력발전소에 면진장치가 적용된다면, 운영기간동안 구조물과 기기들은 동일한 내구성 및 성능이 확보되어야 한다. 본 연구에서는 면진된 원전의 지진에 대한 안전성을 확보하기 위하여 면진구조물의 장기거동을 해석적으로 분석하였다. 경년열화에 의한 면진장치 특성을 분석하였고, 다른 온도환경에서 면진장치의 경년열화에 의한 구조물의 지진응답을 분석하였다. 해석결과에 의하면 면진장치의 경년열화에 의하여 면진구조물의 고유진동수는 증가하였다. 그러나 면진 구조물의 최대가속도와 최대변위는 온도에 따라 크게 변하지 않았다. 면진장치의 열화에 의하여 구조물의 손상은 발생하지 않았지만 목표진동수 영역에서 스펙트럼가속도는 온도에 따라 증가하였다. 따라서 면진구조물에서 면진장치는 온도에 따른 지진응답의 증가를 고려하여 설치 및 제작해야 할 것으로 판단된다.

Probabilistic tsunami hazard analysis (PTHA) is based on the approach of probabilistic seismic hazard analysis (PSHA) which is performed using various seismotectonic models and ground-motion prediction equations. The major difference between PTHA and PSHA is that PTHA requires the wave parameters of tsunami. The wave parameters can be estimated from tsunami propagation analysis. Therefore, a tsunami simulation analysis was conducted for the purpose of evaluating the wave parameters required for the PTHA of Uljin nuclear power plant (NPP) site. The tsunamigenic fault sources in the western part of Japan were chosen for the analysis. The wave heights for 80 rupture scenarios were numerically simulated. The synthetic tsunami waveforms were obtained around the Uljin NPP site. The results show that the wave heights are closely related with the location of the fault sources and the associated potential earthquake magnitudes. These wave parameters can be used as input data for the future PTHA study of the Uljin NPP site.

A methodology to evaluate the seismic performance of interface piping systems that cross the isolation interface in the seismically isolated nuclear power plant (NPP) was developed. The developed methodology was applied to the safety-related interface piping system to demonstrate the seismic performance of the target piping system. Not only the seismic performance for the design level earthquakes but also the performance for the beyond design level earthquakes were evaluated. Two artificial seismic ground input motions which were matched to the design response spectra and two historical earthquake ground motions were used for the seismic analysis of piping system. The preliminary performance evaluation results show that the excessive relative displacements can occur in the seismically isolated piping system. If the input ground motion contained relatively high energy in the low frequency region, we could find that the stress response of the piping system exceed the allowable stress level even though the intensity of the input ground motion is equal to the design level earthquake. The structural responses and seismic performances of piping system were varied sensitively with respect to the intensities and frequency contents of input ground motions. Therefore, for the application of isolation system to NPPs and the verification of the safety of piping system, the seismic performance of the piping system subjected to the earthquake at the target NPP site should be evaluated firstly.