• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1318-1326
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• 2020

Structural modification in the electrical cabinet is investigated by a proposed procedure that comprises of an experimental, analytical and numerical solution. This research emphasizes the linear dynamic analysis of the cabinet that is studied under the seismic excitation to demonstrate the real behavior of the cabinets in NPP. To this end, an actual electric cabinet is experimentally tested using an impact hammer test which reveals the fundamental parameters of the cabinet. The Frequency-domain decomposition (FDD) method is used to extract the dynamic properties of the cabinet from the experiment which is then used for numerical modeling. To validate the dynamic properties of the cabinet an analytical solution is suggested. The calibrated model is analyzed under the floor response obtained from the Connecticut nuclear power plant structure excited by Tabas 1978 (Mw 7.4) earthquake. Eventually, the grouping effect of the cabinets is proposed which represents the influence on the dynamic modification. This grouping of the cabinets is described more sophisticatedly by the theoretical understating, which results in a significant change in the seismic response. Considering the grouping effects will be helpful in the assessment of the real seismic behavior, design, and performance of cabinets.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.1-11
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• 2001

In contrast with conventional fieldbus researches which are focused merely on real-time performance, this study aims to evaluate the real-time performance of the communication system including fault-tolerant mechanisms Maintaining performance in presence of recoverable faults is very important in case that the communication network is applied to a highly reliable system such as next generation Nuclear. Power. Plant (NPP). If the tie characteristics meet the requirements of the system, the faults will be recovered by fieldbus recovery mechanisms and the system will be safe. If the time characteristics can not meet the requirements, the faults in the fieldbus can propagate to the system failure. In this study, for the purpose of investigating the time characteristics of fieldbus, the recoverable faults are classified and then the formulas that represent delays including recovery mechanisms are developed. In order to validate the proposed approach, we have developed a simulation model that represents the Korea Next Generation Reactor (KNGR) NSSS Process Control System (NPCS). The results of the simulation show us the reasonable delay characteristics of the fault cases with recovery mechanisms. Using the simulation results and the system requirements, we also can calculate the failure propagation probability from fieldbus to outer system.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.357-373
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• 2022

The high-temperature gas-cooled reactor pebble-bed module project is the first commercial Generation-IV NPP(Nuclear Power Plant) in China. A new joint is used for the vertical support of RPV(Reactor Pressure Vessel). The steel corbel is integrally embedded into the reactor-cabin wall through eight asymmetrically arranged pre-stressed high-strength bolts, achieving the different path transmission of shear force and moment. The vertical monotonic loading test of two specimens is conducted. The results show that the failure mode of the joint is bolt fracture. There is no prominent yield stage in the whole loading process. The stress of bolts is linearly distributed along the height of corbel at initial loading. As the load increases, the height of neutral axis of bolts gradually decreases. The upper and lower edges of the wall opening contact the corbel plate to restrict the rotation of the corbel. During the loading, the pre-stress of some bolts decreases. The increase of the pre-stress strength ratio of bolts has no noticeable effect on the structure stiffness, but it reduces the ultimate bearing capacity of the joint. A simplified calculation model for the elastic stage of the joint is established, and the estimation results are in good agreement with the experimental results.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3849-3854
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• 2022

The present work aims to optimize the radiation protection efficiency for ion-selective containers used in the liquid treatment for the nuclear power plant (NPP) cooling cycle. Some naturally occurring rocks were examined as filler materials to reduce absorbed dose and equivalent dos received from the radioactive waste container. Thus, the absorbed dose and equivalent dose were simulated at a distance of 1 m from the surface of the radioactive waste container using the Monte Carlo simulation. Both absorbed dose and equivalent dose rate are reduced by raising the filler thickness. The total absorbed dose is reduced from 7.66E-20 to 1.03E-20 Gy, and the equivalent dose is rate reduced from 183.81 to 24.63 µSv/h, raising the filler thickness between 0 and 17 cm, respectively. Also, the filler type significantly affects the equivalent dose rate, where the redorded equivalent dose rates are 24.63, 24.08, 27.63, 33.80, and 36.08 µSv/h for natural rocks basalt-1, basalt-2, basalt-sill, limestone, and rhyolite, respectively. The mentioned results show that the natural rocks, especially a thicker thickness (i.e., 17 cm thickness) of natural rocks basalt-1 and basalt-2, significantly reduce the gamma emissions from the radioactive wastes inside the modified container. Moreover, using an outer cementation concrete wall of 15 cm causes an additional decrease in the equivalent dose rate received from the container where the equivalent dose rate dropped to 6.63 µSv/h.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4679-4683
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• 2022

Environmental radiation monitoring is required to protect from the effects of radiation in industrial fields such as nuclear power plant (NPP) monitoring, and various gamma camera systems are being developed. The purpose of this study is to optimize parameters of a diverging collimator composed of pure tungsten for compactness and lightness through Monte Carlo simulation. We conducted the performance evaluation based on spatial resolution and signal-to-noise ratio for point source and obtained gamma images and profiles. As a result, optimization was determined at a collimator height of 60.0 mm, a hole size of 1.5 mm, and a septal thickness of 1.0 mm. Also, the full-width-at-half-maximum was 3.5 mm and the signal-to-noise ratio was 53.5. This study proposes a compact 45° diverging collimator structure that can quickly and accurately identify the location of the source for radiation monitoring.

• 방사성폐기물학회지
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• 제17권1호
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• pp.15-28
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• 2019

고리 1호기는 원전해체 계획에 따라 영구정지 이후 가능한 한 빠른 시일 내에 원자로냉각재계통의 화학제염을 수행할 계획으로, 계통제염 기술 확보를 위해 한수원에서는 2014년부터 '원전 해체설계를 위한 냉각재계통 및 기기제염 상용기술개발' 연구과제를 통해 화학제염기술을 개발하고 있다. 본 연구를 위해 Lab. 규모 계통제염 공정장치를 제작하였으며, 계통제염 대상의 주요재료인 STS304, 316, 410, Alloy600, SA508을 사용하여 화학제염 공정실험을 수행하였다. 화학제염 공정실험의 목적은 산화-환원공정의 최적시간, 최적제염제 및 공정횟수를 도출하기 위함이다. 화학제염 공정실험은 과망간산-옥살산 기반의 단위공정 및 연속공정 실험, 과망간산+질산-옥살산 기반의 연속공정 실험으로 나누어 수행하였다. 그 결과 단위공정실험을 통해 최적공정 시간인 산화공정 5시간, 환원공정 4시간을 도출하였으며, 연속공정실험을 통해 최적제염제와 공정횟수를 도출하였다. 최적제염제는 산화제의 경우 $200mg{\cdot}L^{-1}$ 과망간산 + $200mg{\cdot}L^{-1}$ 질산이고, 환원제는 $2000mg{\cdot}L^{-1}$ 옥살산이며, 공정횟수는 STS304와 SA508의 경우 2 cycle, Alloy600의 경우 3 cycle 이상 수행하는 것이 적절할 것으로 평가되었다.

• 한국전산유체공학회지
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• 제10권3호
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• pp.9-18
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• 2005

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In order to mitigate hydrogen hazards possibly occurred in the NPP containment, hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) designed in Korea specifies 26 passive autocatalytic recombiners and 10 igniters installed in the containment for the hydrogen mitigation. in this study, the analysis of the hydrogen and steam behavior during a total lose of feed water (TLOFW) accident in the APR1400 containment has been conducted by using the CFD code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released in the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type dampers at the IRWST vents which are operated depending on the pressure difference between inside and outside of the IRWST. it was found that the flaps strongly affects the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and transition from deflagration to detonation (DDT) were evaluated by using Sigma-Lambda criteria. Numerical results indicate the DDT possibility could be heavily reduced in the IRWST compartment when the flaps are installed.

• 한국구조물진단유지관리공학회 논문집
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• 제28권4호
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• pp.48-54
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• 2024

원전 구조물은 냉각수를 사용하기 위해 해안가에 위치하고 있으며, 염해에 의한 철근부식에 노출되어 있다. UAE에 지어지는 원전 구조물의 경우, 해안가의 온도가 높으므로 염화물 이동이 다른 지역에 비하여 빠르게 평가된다. 본 연구에서는 원전 구조물에 사용되어지는 재료와 배합을 이용하여 5,000 psi (35 MPa)설계강도 등급의 시편을 제작하였으며, 온도와 재령을 고려하여 염화물 확산계수를 평가하였다. 재령 28일 및 91일에 강도 평가 및 온도에 따른 확산계수를 평가하여 특성을 분석하였다. 또한 91일 재령 콘크리트에 대하여 20℃~50℃의 범위에서 염화물 확산실험을 수행하였다. 또한 온도에 따른 기울기를 로그함수로 변환하여 활성화에너지를 도출하였으며, 기존의 제안값들과 비교하였다. 제안된 활성화에너지는 온도의존형 염화물 확산계수에 사용하여 합리적인 내구성 설계를 수행할 것으로 평가된다.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2004년도 봄 학술발표회 논문집
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• pp.209-216
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• 2004

Numerical analysis is carried out to identify the appropriateness of the design codes that is available for the tensile design of fastening system at Nuclear Power Plant (NPP) in this study. This study is intended for the cast-in-place anchor that is widely used for the fastening of equipment in Korean NPPs. The microplane model and the elastic-perfectly plastic model are employed for the quasi-brittle material like concrete and for the ductile material like anchor bolt as constitutive model for numerical analysis and smeared crack model is employed for the crack and damage phenomena. The developed numerical model is verified on a basis of the various test data of cast-in-place anchor. The appropriateness of both ACI 349 Code and CCD approach of CEB-FIP Code is evaluated for the tensile design of cast-in-place anchor and it is proved that both design codes give a conservative results compared with real tensile capacity of cast-in-place anchor.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2005년도 춘계 학술발표회 논문집
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• pp.464-470
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• 2005

The numerical analysis is carried out to identify the influence of design factors to shear capacity of cast-in-place (CIP) anchor in ACI 349 Code that is available for the design of fastening system at Nuclear Power Plant (NPP) in this study. The MASA program is used to develop the numerical analysis model and the developed numerical analysis model is verified on a basis of the various test data of CIP anchor. Both $l/d_o$ and $c_1/l$ we considered as design factors. As a result, the variation of $l/d_o$ has no influence on the shear capacity of CIP anchor but $c_1/l$ has a large influence on the shear capacity of CIP anchor, Therefore, it is proved that ACI 349 Code may give a non-conservative results compared with real shear capacity of CIP anchor according to $c_1/l$.