• 한국전산유체공학회:학술대회논문집
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• 2008.03a
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• pp.308-315
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• 2008

극저온 유체를 운반하는 선박의 화물창은 보통 1차 방벽과 2차 방벽으로 구성되어 있다. 1차 방벽에 소량의 극저온 유체의 누출이 생기더라도, 액밀이 되는 2차 방벽에서 추가적인 극저온 유체의 누출을 방지할 수 있기 때문이다. 그러나 2차 방벽에 추가적인 손상이 생길 경우 유리솜으로 만들어진 Flat Joint를 거쳐 선체내벽까지 극저온 유체에 노출될 가능성이 있게 된다. 본 연구의 관심사는 2차 방벽의 손상 정도에 따라, 그리고 누출되는 극저온 유체의 양에 따라, 내측선체에 얼마나 영향이 미치는가를 알아보는 것이다. 이를 위해 본 연구에서는 극저온 유체를 운반하는 선박의 화물창을 구성하는 2차 방벽에 구멍을 뚫어 Flat Joint 사이로 극저온 유체가 흘러 들어가도록 격자를 생성한 후, CFD 소프트웨어를 이용하여 극저온 유체의 누출에 대한 계산을 수행 하였으며 이를 실험 결과와 비교 분석 하였다. 실험과 계산 결과를 살펴보면, 극저온 유체량에 따라 내측 선체에의 피해를 최소화 할 수 있음을 확인하였다.

• 한국전산유체공학회:학술대회논문집
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• 2008.10a
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• pp.308-315
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• 2008

극저온 유체를 운반하는 선박의 화물창은 보통 1차 방벽과 2차 방벽으로 구성되어 있다. 1차 방벽에 소량의 극저온 유체의 누출이 생기더라도, 액밀이 되는 2차 방벽에서 추가적인 극저온 유체의 누출을 방지할 수 있기 때문이다. 그러나 2차 방벽에 추가적인 손상이 생길 경우 유리솜으로 만들어진 Flat Joint를 거쳐 선체내벽까지 극저온 유체에 노출될 가능성이 있게 된다. 본 연구의 관심사는 2차 방벽의 손상 정도에 따라, 그리고 누출되는 극저온 유체의 양에 따라, 내측선체에 얼마나 영향이 미치는가를 알아보는 것이다. 이를 위해 본 연구에서는 극저온 유체를 운반하는 선박의 화물창을 구성하는 2차 방벽에 구멍을 뚫어 Flat Joint 사이로 극저온 유체가 흘러 들어가도록 격자를 생성한 후, CFD 소프트웨어를 이용하여 극저온 유체의 누출에 대한 계산을 수행 하였으며 이를 실험 결과와 비교 분석 하였다. 실험과 계산 결과를 살펴보면, 극저온 유체량에 따라 내측 선체에의 피해를 최소화 할 수 있음을 확인하였다.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.175-184
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• 1996

A nuclide transfer by utilizing mass transfer coefficient and barrier response function defined for each barrier is proposed, by which the final nuclide transfer rate into the sea water can be evaluated. When simple and immediate quantification of the nuclide release is necessary in the conservative aspect, using this kind of approach may be advantageous since each layered barrier can be treated separately from other media in series in the repository system, making it possible to apply separate solutions in succession to other various media. Although one disadvantage is that while flux continuity can be maintained at the interface by using the exit nuclide flux from the first medium as the source flux for the next one, there may be no guarantee for concentration continuity, this problem could be eliminated assuming that there is no boundary resistance to mass transfer across the interface. Mass transfer coefficient can be determined by the assumption that the nuclide concentration gradient at the interface between adjacent barriers remains constant and barrier response function is obtained from an analytical expression for nuclide flow rate out of each barrier in response to a unit impulse into the barrier multiplied by mass transfer coefficient. Total time-dependent nuclide transfer rate from the barrier can then be obtained by convoluting the response function for the barrier with a previously calculated set of time-varying input of nuclide flow rate for the previous barrier.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.73-86
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• 2015

The use of complimentary indicators, other than radiation dose and risk, to assess the safety of radioactive waste disposal has been discussed in a number of publications for providing the reasonable assurance of disposal safety and convincing the public audience. In this study, the radionuclide flux was selected as performance indicator to appraise the performance of engineered barriers and natural barrier in the Wolsong low- and intermediate-level waste disposal facility. Radionuclide flux showing the retention capability by each compartment of the disposal system is independent of assumptions in biosphere model and exposure pathways. The scenario considered as the normal scenario of disposal facility has been divided into intact or degraded silo concrete conditions. In the intact silo concrete, the radionuclide flux has been assessed with respect to the radionuclide retardation performance of each engineered barrier. In the degraded silo concrete, the radionuclide flux has been explored based on the performance degradation of engineered barriers and the relative significance of natural barrier quantitatively. The results can be used to optimally design the near-surface disposal facility being planned as the second project phase. In the future, additional complimentary indicators will be employed for strengthening the safety case for improving the public acceptance of low- and intermediate-level waste disposal facility.

• Journal of Advanced Marine Engineering and Technology
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• v.41 no.4
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• pp.337-344
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• 2017

The demand of liquified natural gas is increasing due to environmental issues. This reason has resulted in increasing the capacity of liquified natural gas cargo tank. The Mark-III type primary barrier directly contacts liquified natural gas. Also, the primary barrier is under various loading conditions such as weight of liquified natural gas and sloshing loads. During a ship operation, various loads can cause fatigue failure. Therefore, the fatigue life prediction should be evaluated to prevent leakage of liquified natural gas. In the present study, the fatigue analysis of insulation system including primary barrier is performed using a finite element model. The fatigue life of primary barrier is carried out using a numerical study. The value of principle stress and the location of maximum principle stress range are calculated, and the fatigue life is evaluated. In addition, the effects on the insulation panel status and the arrangement of knot or corrugation are analyzed by comparing the fatigue life of various models. The insulation system which has best structural performance of primary barrier was selected to ensure structural integrity in fatigue assessment. These results can be used as a design guideline and a fundamental study for the fatigue assessment of primary barrier.

• Tunnel and Underground Space
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• v.14 no.3
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• pp.175-187
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• 2004

고준위방사성폐기물 처분의 경우 심부 암반에 만들어진 처분장에 영구 처분하는 것이 최선의 방안으로 여겨지고 있다. 하지만 지하 심부의 암반에 대한 물리적, 화학적, 역학적, 열적, 수리적 물성과 이들과 핵종 이동의 관계, 처분환경에서의 공학적 방벽 및 암반의 거동이 처분장 안정성 및 안전성에 미치는 영향 등을 파악해야하는 어려움이 따른다. 특히 고준위폐기물 처분의 경우 장기간의 안전성을 고려해야하기 때문에 자연방벽과 공학적 방벽의 시간에 따른 거동변화도 고려하여야 할 필요가 있다. (중략)

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.447-456
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• 1993

A simplified model to assess the performance of engineered barrier for the retardation of radionuclide release in a low- and intermediate-level radioactive waste repository was developed. The model is based on the repository design concept being suggested in Korea, and considers two types of release scenario ; a design-bas release for the design of engineered barrier and a realistic release for the performance assessment. For the numerical illustration, the sample calculations were performed for five radionuclides with different chemical characteristics, and the results were analyzed.

• Proceedings of the Korea Institute of Fire Science and Engineering Conference
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• 2012.04a
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• pp.125-128
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• 2012

국내 가동원전에서는 내화방벽 관통부의 성능이 적절하게 유지되도록 주기점검절차, 데이터베이스 등 발전소 상황에 적합한 관리방안을 수립하여 운영하고 있다. 이 중에서 내화방벽 관통부 유지관리용으로 구축된 데이터베이스는 관통부 유지관리에 필요한 필수정보가 수록되어 있으나 관통부 담당자의 개인 관리도구로 활용되고 있으므로 실효성 향상이 필요하다. 본 논문에서는 이러한 점을 보완하기 위하여 발전소 구조물, 관통부 등의 3차원 형상정보(3-dimensional Plant Model)를 활용하여 시각적인 관리 편의를 향상시키고, 웹 기반(Web-based)의 관통부 정보 구축을 통하여 발전소 수준(Plant-level)에서 관리가 가능한 관통부 관리시스템 개발 사례를 기술하였다.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.36 no.2
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• pp.113-120
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• 2023

This paper introduces a simulation method for the collision between roadblocks and vehicles using LS-DYNA. The need to evaluate the performance of anti-ram barriers to prepare for vehicle impact has increased since vehicle impact threats have been included as a design criterion for nuclear power plants. Anti-ram barriers are typically certified for their performance through collision experiments. However, because Koreas has no performance testing facilities for anti-ram barriers, their performance can only be verified through simulations. LS-DYNA is a specialized program for collision simulation. Various organizations, including NCAC, distributes numerical models that have been validated for their accuracy with collision tests. In this study, we constructed a finite element model of the most critical vehicle barrier module and simulated collision between roadblocks and vehicles. The calculated results were verified by applying the validation criteria for vehicle safety facility collision simulations of NCHRP 179.

• Economic and Environmental Geology
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• v.56 no.5
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• pp.533-545
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• 2023

In this study, an investigation was conducted on the features, events, and processes (FEP) that could impact the long-term safety of the natural barriers constituting high-level radioactive waste geological repositories. The FEP list was developed utilizing the IFEP list 3.0 provided by the Nuclear Energy Agency (NEA) as foundational data, supplemented by geological investigations and research findings from leading countries in this field. A total of 49 FEPs related to the performance of the natural barrier were identified. For each FEP, detailed definitions, classifications, impacts on long-term safety, significance in domestic conditions, and feasibility of quantification were provided. Moreover, based on the compiled FEP list, three scenarios that could affect the long-term safety of the disposal facility were developed. Geological factors affecting the performance of the natural barrier in each scenario were selected and their relationships were visualized. The constructed FEP list and the visualization of interrelated factors in various scenarios are anticipated to provide essential information for selecting and organizing factors that must be considered in the development of mathematical models for quantitatively evaluating the long-term safety of deep geological repositories. In addition, these findings could be effectively utilized in establishing criteria related to the key performance of natural barriers for the confirmation of repository sites.