• Nuclear Engineering and Technology
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• 제15권2호
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• pp.123-134
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• 1983

냉각재 상실 사고시에 격납 용기로부터 열 제거능력의 추정은 격납용기의 기본적인 설계변수인 최고 압력 및 온도의 예측에 있어서 매우 중요하다. 지금까지 격납 용기내의 열전달에 관하여 실험 및 해석적인 방법에 의하여 연구한 결과들을 종합하여 검토해 봄으로서 현재의 수준을 알 수 있었으며, 이들 관계시의 잘못된 사용을 피하고 보다 정확한 예측을 위해서는 더욱 깊은 연구가 필요하다고 생각된다.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2000년도 가을 학술발표회논문집
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• pp.217-221
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• 2000

Nonlinear stress analysis of nuclear containment building wall element is carried out using microscopic material model. The present study mainly focuses on the finite element analysis of the nuclear containment building wall element under biaxial tensile stresses and it evaluates the perfomance of adopted microscopic material model in the membrane energy dominant situation. From the numerical analysis, the adopted material model peforms well and has a good agreement with experiment result. Finally, the result of present study can be severed as a benchmark test when concrete material model is in need of evaluation.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2499-2508
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• 2021

Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance improved when a passive containment cooling heat exchanger (PCCX) was installed in the lower part of the containment building. The OL-PCCS was found to be superior in terms of heat-removal performance. However, in terms of flow instability, the OL-PCCS was more vulnerable than the CL-PCCS. In particular, the possibility of flow instability was higher when the PCCX was installed in the upper part of the containment. Therefore, the installation location of the OL-PCCS should be restricted to minimize flow instability. Conversely, a CL-PCCS can be installed without any positional restriction by adjusting the initial system pressure within the loop, which eliminates flow instability. These results could be used as base data for the thermo-hydraulic evaluation of PCCS in PWR with a large dry concrete-type containment.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2014년도 추계 학술논문 발표대회
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• pp.205-206
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• 2014

It is very important to reduce the construction duration of the Reactor Containment Building (RCB) when considering the more than 50 months on average from concrete placement to completion. The purpose of this study attempts to evaluate the single-stage workability of the system given a change in the height of the setting of RCB exterior wall formwork to be used in nuclear power plant construction. As a result of this study, it is possible height of 3.5m~4m uses formwork when analyzing the construction period and material costs an increase in formwork by concrete lateral pressure, to ensure the workability of the RCB exterior wall formwork. Through this study, I want to provide as basic data for the improvement of workability and RCB shortening the construction period.

• Nuclear Engineering and Technology
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• 제30권1호
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• pp.26-39
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• 1998

The containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5 is modified to incorporate the passive containment cooling models. The correlations are selected from the existing experimental heat transfer correlations to model the natural and mixed convection in annular space between the containment shell and the shield building. The evaporative heat transfer of the water film on the outer shell of the containment is modeled using the correlations derived from the analogy between the heat and mass transfer. The modified code is applied to the Ap600 containment transient analysis for the model verification and the results are compared to the results of GOTHIC calculation done by Westinghouse. Also, d series of parametric sensitivity studies of heat transfer correlations, water film ratio and delay time of the wet cooling on the containment peak pressure and temperature following LOCA are performed for the containment of 1000MWe passive plant, KP1000.

• 방사성폐기물학회지
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• 제18권3호
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• pp.347-354
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• 2020

Site characterization for decommissioning Kori Unit 1 is ongoing in South Korea after 40 years of successful operation. Kori Unit 1's containment building is assumed to be mostly radioactively contaminated, and therefore radiation exposure management and detailed contamination investigation are required for decommissioning and dismantling it safely. In this study, site-specific Derived Concentration Guideline Levels (DCGLs) were derived using the residual radioactivity risk evaluation tool, RESRAD-BUILD code. A conceptual model of containment building for Kori Unit 1 was set up and limited occupational worker building inspection scenario was applied. Depending on the source location, the maximum contribution source and exposure pathway of each radionuclide were analyzed. The contribution of radionuclides to dose and exposure pathways, by source location, is expected to serve as basic data in the assessment criteria of survey areas and classification of impact areas during further decommissioning and decontamination of sites.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.797-802
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• 2014

This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS) for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO) was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.

• 한국건축시공학회지
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• 제15권4호
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• pp.425-431
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• 2015

1978년에 고리 1호기의 건설이 완공된 이래로 원자력 발전 플랜트의 건설 프로젝트는 국내 외로 점차 확대되고 있다. 그러나 일부 원자력 발전 플랜트의 건설 현장에서는 리스크 관리 능력의 부족으로 인하여 공기 지연과 공사비 손실의 문제점들을 가지고 있다. 특히, 원자력 발전 플랜트 내 원자로격납건물의 시공은 타 시공 단계에 비해 긴 공정기간으로 인하여 전문기술과 대규모 자원이 요구됨에 따라 많은 리스크 요인들이 산재될 수 있다. 따라서 원자로격납 건물의 시공에서 예상되는 리스크 요인들을 분석하여 전체 프로젝트의 안정적인 수행 방향을 제시하는 연구가 필요하다. 그러므로 본 연구는 원자로격납건물 시공의 리스크 요인들을 평가하고자 한다. 이를 위하여 본 연구는 36명의 소수 전문가 집단을 대상으로 하는 설문조사방법을 활용하였다. 24개의 리스크 요인들은 공정, 원가, 안전, 품질을 기준으로 분류되었으며, 이에 대한 평가 결과는 계층 분석방법을 활용하여 분석하였다. 이를 바탕으로 각 기준별로 분류된 리스크 요인들은 중요도와 우선순위를 산정하고 원자력 발전 플랜트의 시공 리스크 요인을 분석하는데 계층분석 방법의 적용성을 확인하였다. 본 연구의 결과는 원자로격납건물의 시공 단계에서 리스크 관리를 위한 기초 자료로 활용될 수 있을 것이다.

• Structural Engineering and Mechanics
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• 제53권1호
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• pp.73-87
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• 2015

In case of aircraft impact on nuclear containment structures, the initial kinetic energy of the aircraft is transferred and absorbed by the outer containment, may causing either complete or partial failure of containment structure. In the present study safety analysis of BWR Mark III type containment has been performed. The total height of containment is 67 m. It has a circular wall with monolithic dome of 21m diameter. Crash analysis has been performed for fighter jet Phantom F4. A normal hit at the crown of containment dome has been considered. Numerical simulations have been carried out using finite element code ABAQUS/Explicit. Concrete Damage Plasticity model have been incorporated to simulate the behaviour of concrete at high strain rate, while Johnson-Cook elasto-visco model of ductile metals have been used for steel reinforcement. Maximum deformation in the containment building has reported as 33.35 mm against crash of Phantom F4. Deformations in concrete and reinforcements have been localised to the impact region. Moreover, no significant global damage has been observed in structure. It may be concluded from the present study that at higher velocity of aircraft perforation of the structure may happen.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2004년도 춘계 학술발표회 제16권1호
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• pp.758-761
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• 2004

Prestressing tendons of the nuclear containment building dome are arranged in a non-axisymmetric manner. However, simple axisymmetric modeling of the containment building is often employed to estimate the structural behavior for, e.g. the ultimate pressure, which requires the axisymmetric approximation of the actual tendon arrangements of the dome. A procedure is proposed that can devise the actual 3-dimensional tendon stiffness and prestressing effect into the axisymmetric model. A numerical example of the CANDU type is presented to verify the procedure and to estimate the amount of approximation.