• The Journal of the Korea Contents Association
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• v.16 no.12
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• pp.373-382
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• 2016

Carbon dioxide Capture and Storage/Sequestration (CCS) technology has attracted attention as an ideal method for most carbon dioxide reduction needs. When the collected carbon dioxide is transported to storage via pipelines, the direct transport is made if the storage is close, otherwise it can also be transported via an intermediate storage hub. Determining the number and the location of the intermediate storage hubs is an important problem. A decision-making algorithm using a mathematical model for solving the problem requires considerably more variables and constraints to describe the multi-objective decision, but the computational complexity of the problem increases and it also does not guarantee the optimality. This research proposes an algorithm to determine the location and the number of the intermediate storage hub and develop a simulator for the connection network of the carbon dioxide emission site. The simulator also provides the course of transportation of the carbon dioxide. As a case study, this model is applied to Korea.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.11a
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• pp.246-247
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• 2006

• Solar Energy
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• v.17 no.3
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• pp.13-21
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• 1997

In this study, the theoretical equation of thermal storage efficiency was established to applied long term hot water storage system. The, effective thermal diffusivity and storage efficiency were, measured through the experiment to predict the degree of mixture in thermal storage tank. The effective thermal diffusivity was inversely preportional to the storage efficiency. The most effective storage efficiency was obtained under condition of low flow rate and using the perforated distributor.

• Nuclear industry
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• v.5 no.2 s.24
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• pp.50-53
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• 1985

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2010.05a
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• pp.353-354
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• 2010

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2008.05a
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• pp.68-69
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• 2008

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.92-100
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• 2015

For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.596-601
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• 1995

신뢰도 분석 방법을 이용하여 습식 사용후핵연료 중간저장시설의 냉각계통에 대한 최적 설계조건을 도출하기 위한 연구를 수행하였다. 먼저 고장수목 분석을 통한 설계 취약점을 평가하여 21개의 설계대안을 도출하였고, 최종적으로 설계대안에 대한 건설비 용, 계통신뢰도 분석 및 확률론적 안전기준을 고려한 비용효과 분석을 실시하였다. 설계 대안들 중에서 100% 루프 다중설계, 루프당 한 개의 펌프 사용, 안전등급 부여 및 주 루프에서 정화계통이 분리된 경우가 최적설계안으로 나타났다. 여기서 적용된 방법론은 유사시설의 최적설계에 유용하게 응용될 수 있을 것으로 사료된다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.391-402
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• 2017

The Korea Radioactive Waste Agency (KORAD) has developed a concrete cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. During long-term storage of spent nuclear fuel in concrete casks kept in dry conditions, the integrity of the concrete cask and spent nuclear fuel must be maintained. In addition, the radiation dose rate must not exceed the storage facility's design standards. A suitable shielding design for radiation protection must be in place for the dry storage facilities of spent nuclear fuel under normal and accident conditions. Evaluation results show that the appropriate distance to the annual dose rate of 0.25 mSv for ordinary citizens is approximately 230 m. For a $2{\times}10$ arrangement within storage facilities, rollover accidents are assumed to have occurred while transferring one additional storage cask, with the bottom of the cask facing the controlled area boundary. The dose rates of 12.81 and 1.28 mSv were calculated at 100 m and 230 m from the outermost cask in the $2{\times}10$ arrangement. Therefore, a spent nuclear fuel concrete cask and storage facilities maintain radiological safety if the distance to the appropriately assessed controlled area boundary is ensured. In the future, the results of this study will be useful for the design and operation of nuclear power plant on-site storage or intermediate storage facilities based on the spent fuel management strategy.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.197-201
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• 2012

Annual dose at the boundary of the interim storage facility at normal condition was calculated to estimate the site area of the facility of PWR spent nuclear fuel. In this work, source term was generated by ORIGEN-ARP for 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facilities and radiation shielding evaluations were conducted by MCNP code depending on the storage capacity. In the case of the centralized storage system, the required site area was found to have the radius of more than 700 m.