• 한국정밀공학회지
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• 제28권2호
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• pp.233-238
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• 2011

In piping design of nuclear power plant facilities, the load stress according to self-weight is important for design values in test run(shutdown and starting). But sometimes it needs more studies, such as seismic analysis of an earthquake of power plant area and fatigue life and stress of thermal expansion and anchor displacement in operating run. In this paper, seismic evaluations were performed to nuclear piping system of Shin-Kori NO. 3&4 being built in Pusan lately. Results of seismic analysis are evaluated on basis of KEPIC MN code. The structural integrity on RCB piping system was proved.

• 한국전산구조공학회논문집
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• 제23권2호
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• pp.235-243
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• 2010

본 논문에서는 원자로 격납건물의 3차원 해석을 수행할 수 있는 구조해석 시스템을 구축하여 제시하였다. 구조해석 시스템은 고성능 평판 및 쉘 유한요소를 요소 라이브러리로 추가하였고, 비부착식 텐던과 부착식 텐던의 거동을 정확하게 모사할 수 있는 모델링방법을 포함하고 있다. 이러한 기능을 프로그래밍하고 범용 구조해석프로그램 DIANA에 접목시켜 원자로 격납건물의 비선형해석은 물론이고 내압능력 평가가 가능하다. 본 논문에서 제안한 3차원 구조해석 시스템의 신뢰성을 확인하기 위해 중수로형 원자로 격납건물의 구조해석을 수행하여 다른 기관에서 수행한 축대칭 구조해석 결과와 비교분석하였다.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2696-2707
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• 2021

The purpose of this study is to investigate the efficiency of various structural modeling schemes for evaluating seismic performances and fragility of the reactor containment building (RCB) structure in the advanced power reactor 1400 (APR1400) nuclear power plant (NPP). Four structural modeling schemes, i.e. lumped-mass stick model (LMSM), solid-based finite element model (Solid FEM), multi-layer shell model (MLSM), and beam-truss model (BTM), are developed to simulate the seismic behaviors of the containment structure. A full three-dimensional finite element model (full 3D FEM) is additionally constructed to verify the previous numerical models. A set of input ground motions with response spectra matching to the US NRC 1.60 design spectrum is generated to perform linear and nonlinear time-history analyses. Floor response spectra (FRS) and floor displacements are obtained at the different elevations of the structure since they are critical outputs for evaluating the seismic vulnerability of RCB and secondary components. The results show that the difference in seismic responses between linear and nonlinear analyses gets larger as an earthquake intensity increases. It is observed that the linear analysis underestimates floor displacements while it overestimates floor accelerations. Moreover, a systematic assessment of the capability and efficiency of each structural model is presented thoroughly. MLSM can be an alternative approach to a full 3D FEM, which is complicated in modeling and extremely time-consuming in dynamic analyses. Specifically, BTM is recommended as the optimal model for evaluating the nonlinear seismic performance of NPP structures. Thereafter, linear and nonlinear BTM are employed in a series of time-history analyses to develop fragility curves of RCB for different damage states. It is shown that the linear analysis underestimates the probability of damage of RCB at a given earthquake intensity when compared to the nonlinear analysis. The nonlinear analysis approach is highly suggested for assessing the vulnerability of NPP structures.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.11-25
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• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.

• 대한기계학회논문집B
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• 제23권10호
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• pp.1274-1284
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• 1999

The Regulatory Guide 1.82 recommends an analysis of hydraulic performance for sump of ECCS (Emergency Core Cooing System) when LOCA(Loss of Coolant Accident) occurs in a nuclear power plant. The present study deals with 3-dimensional, unsteady, turbulent and two-phase flow simulation to examine the behavior of mixture of reactor coolant and debris near the floor of containment building in conjunction with appropriate assumptions. The dispersed solid model has been adjusted to the interfacial momentum transfer between reactor coolant and debris. According to the results, the counterclockwiserecirculation zone had been formed in the region between sump and connection aisle about 376 second after LOCA occurs. The debris thickness accumulated on a sump screen periodically increases or decreases up to 2000 second, afterwards its peak decreases.

• 한국건축시공학회지
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• 제15권6호
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• pp.653-660
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• 2015

Technologies for reducing construction duration are key factors in nuclear power plant construction projects, as a reduction in construction duration at the construction phase leads to a reduction in construction cost and an increase in profits through the early operation of the nuclear power plant. To analyze the constructability of the height of single-layer placement of formwork for the Reactor Containment Building (RCB) exterior wall through lateral pressure according to the height of concrete placement, the deformation criteria for formwork, and a new form design, 'MIDAS GEN (hereinafter referred to as MIDAS)' is used in this study. The cost and workload of formwork are derived according to the unit of height of the RCB exterior wall. Based on the result, it was found that the higher the RCB exterior wall, the higher the material cost, and the less the construction duration and the less the total number of formwork layers. Based on this result, it is believed that the material cost and the construction duration can be appropriately determined according to the formwork height.

• 대한토목학회논문집
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• 제32권3A호
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• pp.183-188
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• 2012

국내 일부 가동 중 원전의 원자로건물에 부착식 텐던이 시공되어 있고, 이들에 대한 긴장력 평가는 원자로건물의 구조 건전성 평가 시 매우 중요하다. 따라서, 본 논문에서는 기존의 간접적인 부착식 텐던의 긴장력 평가방법을 개선하기 위하여 개발된 SI 기술과 충격신호 분석기술을 이용하여 실제 원자로건물에 매입된 부착식 텐던을 대상으로 긴장력을 평가하였다. 이를 위해 원자로건물에서 발생하는 변형률과 부착식 텐던에서 발생하는 응력파속도를 계측하였다. 이들을 통해 부착식 텐던의 긴장력을 평가한 결과, SI 기술과 충격신호 분석기술 모두 높은 신뢰성 있는 결과를 제시하였고, 기존의 이론적인 접근 방법에 의한 결과와도 매우 유사한 경향을 제시함으로써 본 연구진에서 개발한 부착식 텐던의 긴장력 평가방법이 매우 유용함을 확인할 수 있었다.

• 한국전산구조공학회논문집
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• 제34권6호
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• pp.385-392
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• 2021

강진 시 원자력발전시설의 비선형 응답이 중요하기 때문에 이 시설의 내진성능에 대한 관심이 증가하였다. 이 연구에서는 원자력 발전소 철근콘크리트 전단벽의 유한요소해석을 위한 재료모델의 적절한 변수를 제시하였다: 최대인장강도, 팽창각, 손상계수. 이를 위해 상용 유한요소 해석프로그램인 ABAQUS를 사용하여 낮은 형상비를 가진 철근콘크리트 전단벽의 비선형 거동과 전단 파괴모드에 대한 이 주요 변수의 효과에 대한 연구를 수행하였다. 연구결과에 기반하여 비선형 시간이력해석을 통해 강진 하의 원자로건물의 비선형 응답을 평가하였다.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2748-2754
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• 2022

Because existing reactant coolant system (RCS) leakage detection mechanisms are insensitive to small leaks, a real-time, direct detection system with a detection threshold below 0.5 gpm·hr^-1 was studied. A beta-ray detection system using a silicon detector with good energy resolution for beta rays and a low gamma-ray response was proposed. The detection performance in the leakage condition was evaluated through experiments and simulations. The concentration of ¹⁶N in the coolant corresponding to a coolant leakage of 0.5 gpm was calculated using the analytic method and ORIGEN-ARP. Based on the concentration of ¹⁶N and the measurement of the silicon detector with ⁹⁰Sr/⁹⁰Y, the beta-ray count rate was estimated using MCNPX. To evaluate the effect of gamma rays inside the containment building, the signal-to-noise ratio (SNR) was calculated. To evaluate the count rate ratio, the radiation field inside the containment building was simulated using MCNPX, and response evaluation experiments were performed using beta and gamma rays on the silicon detector. The expected beta-ray count rate at 0.5 gpm leakage was 7.26 × 10⁵ counts/sec, and the signal-to-background count rate ratio exceeded 88 for a transport time of 10 s, demonstrating its suitability for operation inside a reactor containment building.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1567-1576
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• 2023

The reactor containment building (RCB) in nuclear power plants (NPPs) plays an important role in protecting the reactor systems from external loads as well as preventing radioactive leaking. As we witnessed the nuclear disaster at Fukushima Daiichi (Japan) in 2011, the earthquake is one of the major threats to NPPs. The purpose of this study is to evaluate effects of concrete material models and presstressing forces on the seismic performance evaluation of RCB in NPPs. A typical RCB designed in Korea is employed for a case study. Detailed three-dimensional nonlinear finite element models of RCB are developed in ANSYS. A series of pushover analyses are then performed to obtain the pushover curves of RCB. Different capacity curves are compared to recognize the influence of different material models on the nonlinear behavior of RCB. Additionally, the effects of prestressing forces on the seismic performances of the structure are also investigated. Moreover, a set of damage states corresponding to damage evolutions of the structures is proposed in this study.