• Journal of Radiation Protection and Research
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• 제28권4호
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• pp.291-298
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• 2003

The KN-12 transport cask has been designed to transport 12 PWR spent nuclear fuel assemblies and to comply with the regulatory requirements for a Type B(U) package. The containment boundary of the cask is defined by a cask body, a cask lid, lid bolts with nuts, O-ring seals and a bolted closure lid. The containment vessel for the cask consists of a forged thick-walled carbon steel cylindrical body with an integrally-welded carbon steel bottom and is closed by a lid made of stainless steel, which is fastened to the cask body by lid bolts with nuts and sealed by double elastomer O-rings. In the cask lid an opening is closed by a plug with an O-ring seal and covered by the bolted closure lid sealed with an O-ring. The cask must maintain a radioactivity release rate of not more than the regulatory limit for normal transport conditions and for hypothetical accident conditions, as required by the related regulations. The containment requirements of the cask are satisfied by maintaining a maximum air reference leak rate of $2.7{\times}10^{-4}ref.cm^3s^{-1}$ or a helium leak rate of $3.3{\times}10^{-4}cm^3s^{-1}$ for normal transport conditions and for hypothetical accident conditions.

• Nuclear Engineering and Technology
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• 제33권5호
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• pp.468-482
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• 2001

As a process of Direct Containment Heating (DCH) issue resolution for Korean Standard Nuclear Power Plants (KSNPs), a containment load/strength assessment with two different approaches, the probabilistic and the deterministic, was performed with all plant-specific and phenomena-specific data. In case of the probabilistic approach, the framework developed to support the Zion DCH study, Two-Cell Equilibrium (TCE) coupled with Latin Hypercubic Sampling (LHS), provided a very efficient tool to resolve DCH issue. In case of the deterministic approach, the evaluation methodology using the sophisticated mechanistic computer code, CONTAIN 2.0 was developed, based on findings from DCH-related experiments or analyses. For three bounding scenarios designated as Scenarios V, Va, and VI, the calculation results of TCE/LHS and CONTAIN 2.0 with the conservatism or typical estimation for uncertain parameters, showed that the containment failure resulted from DCH loads was not likely to occur. To verify that these two approaches might be conservative , the containment loads resulting from typical high-pressure accident scenarios (SBO and SBLOCA) for KSNPs were also predicted. The CONTAIN 2.0 calculations with boundary and initial conditions from the MAAP4 predictions, including the sensitivity calculations for DCH phenomenological parameters, have confirmed that the predicted containment pressure and temperature were much below those from these two approaches, and, therefore, DCH issue for KSNPS might be not a problem.

• 한국지반신소재학회논문집
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• 제7권3호
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• pp.1-7
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• 2008

본 연구에서 검토한 지오텍스타일 컨테이너와 튜브를 활용한 수로구조물은 UAE에서 수행한 프로젝트로 수심이 4.0m 이하인 경우 지오텍스타일 컨테이너를 저개식 바지로부터 해저바닥에 착지시키는 방법으로 시공하고 그 윗부분은 지오텍스타일 튜브로 시공을 수행하였다. 따라서, 본 연구에서는 지오텍스타일 컨테이너와 튜브를 활용하여 해상의 발전소 방파제 수로구조물 축조 시 고려된 구조물의 개별 및 다단구조물의 안정성, 지오텍스타일 재질 선택에 관련한 설계인자를 분석하였다. 또한, 방파제 구조물이 위치한 현장이 발전소의 용수의 유입과 유출이 발생하는 구역으로 발전소 유출입수의 온도가 방파제 구조물 내부의 온도변화에 미치는 영향에 대하여 분석하였다.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.236-245
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• 2016

In a reactor coolant system of a nuclear power plant (NPP), an overpressure protection system keeps pressure in the loop within 110% of design pressure. However if the system does not work properly, pressure in the loop could elevate hugely in a short time. It would be seriously disastrous if a weak point in the pressure boundary component bursts and releases radioactive material within the containment; and it may lead to a leak outside the containment. In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM) analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization. It revealed which component was the weakest and which component had the highest margin to bursting failure. This information is valuable in severe accident progression prediction. The burst pressures of APR-1400, AP1000 and VVER-1000 reactor coolant systems were evaluated and compared to give relative margins of safety.

• 한국지진공학회논문집
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• 제17권6호
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• pp.245-255
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• 2013

According to natural frequency of soil, characteristics of earthquake responses of an isolated containment building in nuclear power plants are examined. For this, earthquake response analysis of seismically isolated containment buildings in nuclear power plants is carried out by strictly considering soil-structure interactions. The structure and near-field soil are modeled by the finite element method while far-field soil by consistent transmitting boundary. The equation of motion of a soil-structure interaction system under incident seismic wave is derived. The derived equations of motion are solved to carry out earthquake analysis of a seismically isolated soil-structure system. Generally, the results of this analysis show that seismic isolation significantly reduces the responses of the soil-structure system. However, if the natural frequency of the soil is similar to that of the soil-structure system, the responses of the containment buildings in nuclear power plants rather increases due to interactions in the system.

• 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.

• 한국전산구조공학회논문집
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• 제34권3호
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• pp.167-174
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• 2021

최근 국내 원자력발전소의 격납건물 벽체와 Containment Liner Plate(CLP) 사이에서 다양한 크기의 공극이 발견됨에 따라 원전 격납건물의 보수를 위해 내부 공극의 분포와 크기를 정밀하게 평가할 수 있는 진단기법의 개발이 요구되고 있다. 이에 따라 이 연구에서는 격납건물 벽체에서의 탄성파 전파거동을 계산하는 2차원 유한요소해석 기법을 제시한다. 격납건물 벽체를 기반으로 해석영역을 구성하고 경계면에서의 반사파를 제거하기 위해 수치적 파동흡수 경계층인 perfectly matched layer를 도입하였다. Galerkin 기반 혼합유한요소법을 이용해 2차원 유한영역에서 탄성파 파동방정식의 해를 구하여 충격하중에 대한 격납건물 벽체의 변위와 응력을 계산하였다. 제시한 수치적 기법을 이용하여 격납건물 콘크리트 벽체의 CLP 부착 유무와 공동의 위치 및 크기 변화에 따른 탄성파 전파거동을 살펴보았다. 이 연구의 결과는 원전 격납건물 내부의 공동을 진단하는 탄성파 전체파형 역해석 기법 개발에 활용될 수 있다.

• Nuclear Engineering and Technology
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• 제47권5호
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• pp.567-578
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• 2015

In a nuclear reactor containment, wall condensation forms with noncondensable gases and their accumulation near the condensate film leads to a significant reduction in heat transfer. In the framework of nuclear reactor safety, the film condensation in the presence of noncondensable gases is of high relevance with regards to safety concerns as it is closely associated with peak pressure predictions for containment integrity and the performance of components installed for containment cooling in accident conditions. In the present study, CUPID code, which has been developed by KAERI for the analysis of transient two-phase flows in nuclear reactor components, is improved for simulating film condensation in the presence of noncondensable gases. In order to evaluate the condensate heat transfer accurately in a large system using the two-fluid model, a mass diffusion model, a liquid film model, and a wall film condensation model were implemented into CUPID. For the condensation simulation, a wall function approach with a heat/mass transfer analogy was applied in order to save computational time without considerable refinement for the boundary layer. This paper presents the implemented wall film condensation model, and then introduces the simulation result using the improved CUPID for a conceptual condensation problem in a large system.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2006년도 정기 학술대회 논문집
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• pp.639-646
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• 2006

A nonlinear finite element analysis is carried out to predict the ultimate internal pressure and failure mechanism of a 1/4 scale prestressed concrete containment vessel(PCCV) model using the commercial code ABAQUS. Therefore, this paper is mainly focused to compare the influence of concrete material model, tension stiffening parameter, uplift phenomenon and basemat. From the analysis results, nonlinear behavior of the PCCV showed a substantially different aspects in accordance with the nonlinear material model for the concrete as well as tension stiffening parameter. The boundary conditions beneath the basemat are considered to be a fixed condition and a nonlinear spring element to compare the influence of the uplift. The finite element analysis is considered with and without a basemat to find out the influence of the basemant itself. From the analysis results, the nonlinear behavior of the PCCV is entirely similar for the two cases.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2524-2533
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• 2024

When analyzing containment thermal-hydraulics, computational fluid dynamics (CFD) is a powerful tool because multi-dimensional and local analysis is required for some accident scenarios. According to the previous study, neglecting steam bulk condensation in the CFD analysis leads to a significant error in boundary layer profiles. Validating the condensation model requires the experimental data near the condensing surface, however, available boundary layer data is quite limited. It is also important to confirm whether the heat and mass transfer analogy (HMTA) is still valid in the presence of bulk condensation. In this study, the boundary layer measurements on the vertical condensing surface in the presence of air were performed with the rectangular channel facility WINCS, which was designed to measure the velocity, temperature, and concentration boundary layers. We set the laminar flow condition and varied the Richardson number (1.0-23) and the steam volume fraction (0.35-0.57). The experimental results were used to validate CFD analysis and HMTA models. For the former, we implemented a bulk condensation model assuming local thermal equilibrium into the CFD code and confirmed its validity. For the latter, we validated the HMTA-based correlations, confirming that the mixed convection correlation reasonably predicted the sum of wall and bulk condensation rates.