• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2004년도 추계 학술발표회 제16권2호
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• pp.381-384
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• 2004

Delaying and objection for the construction of storage spent-fuel disposal has prompted to consider expanding on-site storage of spent reactor fuel since it can eliminate the need for costly and difficult shipping and control of the spent fuel completely under the direction of the owner-utility. The dry storage unit developed in Canada can accommodate Korea heavy water reactor fuel elements and become a candidate for the Korean market. In this paper, finite element analyses were carried out in order to investigate the structural behavior of the nuclear spent fuel dry storage system, which is subjected to impact loads such as collision of a truck load and dropping of flask under the irregular operation.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.533.1-533
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• 2001

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1127-1133
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• 2021

A Spent Fuel Pool (SFP) is designed to store spent fuel assemblies in the pool. And, a SFP cooling and cleanup system cools the SFP coolant through a heat exchanger which exchanges heat with component cooling water. If the cooling system fails or interfacing pipe (e.g., suction or discharge pipe) breaks, the cooling function may be lost, probably leading to fuel damage. In order to prevent such an incident, it is required to properly cool the spent fuel assemblies in the SFP by either recovering the cooling system or injecting water into the SFP. Probabilistic safety assessment (PSA) is a good tool to assess the SFP risk when an initiating event for the SFP occurs. Since PSA has been focused on reactor-side so far, it is required to study on the framework of PSA approach for SFP and identify the key factors in terms of fuel damage frequency (FDF) through a case study. In this study, therefore, a case study of SFP-PSA on the basis of design information of APR-1400 has been conducted quantitatively, and several sensitivity analyses have been conducted to understand the impact of the key factors on FDF.

• Nuclear Engineering and Technology
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• 제43권1호
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• pp.75-82
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• 2011

The neutron multiplication factor in spent fuel storage racks, in which a stainless steel plate encloses a fuel assembly, was evaluated according to the variation of distance between the fuel assembly and stainless steel plate, as well as the pitch. The stainless steel plate position with the lowest multiplication factor on each pitch consistently appeared as 6mm or 9mm away from the outmost surface of the fuel assembly. Because the stainless steel plate has a thermal neutron absorption cross section, its ability to absorb neutrons can work best only if it is installed at the position where thermal neutrons can be gathered most easily. Therefore, the stainless steel plate position should not be too close or too far away from the fuel assembly, but it should be kept a pertinent distance from the fuel assembly.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1109-1119
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• 2016

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4597-4606
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• 2023

Thermal analysis of the CANDU spent fuel dry storage canister is very important to ensure the integrity of the spent fuel. The analyses have been conducted using a conservative approach, with a particular focus on the peak cladding temperature (PCT) of the fuel rods in the canister. In this study, we have performed a realistic thermal analysis using a computational fluid dynamics (CFD) code. The canister contains 9 fuel bundle baskets. A detailed analysis of even a single basket requires significant computational resources. To overcome this challenge, we replaced each basket with an equivalent heat conductor (EHC), of which effective thermal conductivity (ETC) is developed from the results of detailed CFD calculations of a fuel bundle basket. Then, we investigated the effects of some conservative models, ultimately aiming at a realistic analysis. The results revealed: (i) The influence of convective heat transfer in the basket cannot be ignored, but it's less significant than expected. (ii) Modeling of the lifting rod is crucial, as it plays a decisive role in axial heat transfer at the center of the canister and significantly reduces the PCT. (iii) Convection within the canister is very important, as it not only reduces the PCT but also shifts its location upwards.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1439-1447
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• 2023

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2-4 β_eff for 20 years, and achieving a burnup of 9.9-10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.127-128
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• 2018

• Journal of Radiation Protection and Research
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• 제25권2호
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• pp.89-96
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• 2000

원자력 정책에서 안전성과 운영실적, 환경보전, 경제성 등은 매우 중요한 인자이다. 핵주기의 선택은 에너지정책, 연료의 다양성, 공급의 안정과 관련된 모든 사회적, 환경적 영향에 있어 매우 중요하다. 특히, 원전의 고준위 방사성폐기물인 사용후 핵연료 관리는 높은 방사선준위 뿐만 아니라 장기적인 관리기간이 소요되는 어려운 사업이다. 본 연구는 사용후 핵연료 관리방안인 재처리와 직접 처분의 비용분석, 안전성, 대국민 용인 측면을 살펴보았다. 직접 처분이 재처리에 비해 약 7%정도의 경제성이 있고, 직접 처분의 사용후 핵연료는 재처리폐기물보다 높은 위험도를 갖는다. 대국민 용인측면에서는 두 가지 처리방법 모두 찬성하지 않는다. 결론적으로, 사용후 핵연료관리는 모든 사회/환경적 영향과 경제성을 고려한 핵주기 정책과 병행하여 지속적인 기술개발을 통한 안전성확보가 필요하다.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2004년도 ICCAS
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• pp.840-845
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• 2004

The remote operation of the Advanced Spent Fuel Conditioning Process (ACP) is analyzed by using the 3D graphic simulation tools. Since the spent nuclear fuel, which is a high radioactive material, is processed in the ACP, the ACP equipment is operated in intense radiation fields as well as in a high temperature. Thus, the equipment is operated in a remote manner and should be designed with consideration for the remote handling and maintenance. Also suitable remote handling technology needs to be developed along with the design of the process concepts. For this we developed a graphic simulator, which provides the capability of verifying the remote operability of the ACP without the fabrication of the process equipment. In other words, by applying virtual reality to the remote maintenance operation, a remote operation task can be simulated in the graphic simulator, not in the real environment. The graphic simulator will substantially reduce the cost of the development of the remote handling and maintenance procedure as well as the process equipment, while at the same time developing a remote maintenance concept that is more reliable, easier to implement, and easier to understand.