• Journal of Radiation Industry
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• v.17 no.4
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• pp.519-532
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• 2023

Spent nuclear fuel(SNF) is stored in nuclear power plants for a certain period of time and then transported to an interim storage facility. After that, SNF is finally repackaged in a disposal canister at an encapsulation plant for final disposal. Finland and Sweden, leading countries in SNF disposal technology, have already completed designing of spent fuel encapsulation plant. In particular, the encapsulation plant construction in Finland is near completion. When it comes to South Korea, as the amount of SNF production and disposal plan is different from those in Finland and Sweden, it is difficult to apply the concepts of these contries as is. Therefore, it is necessary to establish the spent fuel repackaging concept and to derive each operating and repackaging procedures by considering annual disposal plan of South Korea. The results of this study is expected to be used to establish the concept of optimized encapsulation plant through further research.

• Nuclear Engineering and Technology
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• v.48 no.5
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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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• v.55 no.12
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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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• v.55 no.4
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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.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.127-128
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• 2018

• Journal of Radiation Protection and Research
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• v.25 no.2
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• pp.89-96
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• 2000

Nuclear fuel cycle choices and costs are important in considering energy policies, fuel diversity, security of supply and associated social and environmental impacts. Particularly, the nuclear spent fuel is very important in view of high activity and the need of long term management. This study focuses on the comparison of reprocessing and direct disposal of nuclear spent fuel in terms of cost, safety and public acceptability. The results of the study show that the direct disposal is about 7% more economical than the reprocessing. In terms of safety, the results show that the risk of vitrified HLW (high-level radioactive waste) is less than directly disposed spent fuel. For the public acceptability, both of the methods are not well understood and therefore they are not accepted. In conclusion, it is necessary to guarantee the safety of the both spent fuel processing methods through continuous development of associated technology and to have a fuel cycle policy which should consider not only the economics but also social and environmental impacts.

• 제어로봇시스템학회:학술대회논문집
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• 2004.08a
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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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.269-277
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• 2010

Various multilateral approaches to the nuclear fuel cycle have been proposed in order to suppress the expansion of sensitive fuel cycle technology. In order to prepare for the future multilaterallization of the nuclear fuel cycle, existing multilateral spent fuel management programs are analyzed. A trial multilateralization of a domestic R&D facility for the back end of the nuclear fuel cycle is proposed and its challenges, possibilities and implementation strategy are discussed.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.20 no.2
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• pp.157-164
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• 2007

The prediction of the temperature distribution change of the spent nuclear fuel disposal canister and its surrounding structures (bentonite buffer, granitic rock etc.) due to the spent fuel heat is very important for the design of the 500m deep granitic repository for the spent nuclear fuel disposal canister (about 10,000 years long) deposition. In this study, the temperature distribution change of the composite structure which comprises the canister, the bentonite buffer, the deposition tunnel due to the spent fuel heat is computed using the numerical analysis method. Specially, the temperature distribution change of the composite structure is analysed as the deposition time elapses up to m years. The analysis result shows that the temperature of each part of the repository increases slowly in different way but the latest part temperature increases slowly up to 150 years and thereafter decreases slowly.

• Analytical Science and Technology
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• v.18 no.4
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• pp.298-308
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• 2005

The methods for determining C-14 and tritium contents in the spent nuclear fuel sample were developed. The carbon-14($^{14}CO_2$) released during the dissolution of the spent fuel sample and $CaCO_3$ ($CO_2$ carrier) with 8 M $HNO_3$ at $90^{\circ}C$ was collected in trap containing 1.5 M NaOH. The volatile radioactive iodine evolved when the spent fuel was dissolved, was trapped on to Ag-silicagel (Ag-impregnated silicagel) adsorbent in column which is connected to two NaOH traps. The solutions which contain tritium as HTO after fuel dissolution were decontaminated by deionization with a mixture of cation and anion exchange resins and inorganic ionexchangers. The amount of C-14 in the trap solutions and the HTO concentration in the resulting deionization water were then determined by liquid scintillation counting.