• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.225-235
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• 2005

This paper outlined the status of the development of Korean Reference Disposal (KRS­1) system for high-level radioactive wastes. The repository concept was based on the engineering barrier system which KAERI has developed through a long-term research and development program. The design requirements were prepared for the conceptual design of the repository. The amount of PWR and CANDU spent fuels were projected with the current nuclear power plan. The disposal rates of PWR and CANDU spent fuels were analyzed. The reference geologic characteristics including classification of fracture zones were set for the KRS. The disposal concepts and the layout of the repository were described.

• 융복합기술연구소 논문집
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• 제4권2호
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• pp.47-50
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• 2014

Voloxidation is a process for converting $UO_2$ into $U_3O_8$ while removing some volatile products in spent fuels (SF). Various oxidative gas conditions including air and mixture of Ar and $O_2$ could be adopted for the process. The gas flows into a reactor under high temperature ($>500^{\circ}C$) and components of SF are reacted with the gas. SF is composed of various components such as actinides, lanthanides, and alkali metals. Therefore, it is of significance to understand their behavior during the reactions for process development. However, due to the limit of available experiments, phase diagram analysis should be preceded. TPP diagram is constructed with respect to temperature-pressure-pressure. It shows a stable phase depending on partial pressures of gas components as well as temperature. In this work, we investigated TPP diagrams for actinides, lanthanides and other oxides to determine stable oxide forms under different gas conditions. The results would be used to set up a material balance under a pyroprocessing scheme of SF and compare the gas conditions for the optimization of fission products removal.

• Nuclear Engineering and Technology
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• 제40권7호
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• pp.581-592
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• 2008

In this study, a conceptual pyroprocess flowsheet has been devised by combining several dry-type unit processes; its applicability as an alternative fuel cycle technology was analyzed. A key point in the evaluation of its applicability to the fuel cycle was the recovery yield of fissile materials from spent fuels as well as the proliferation resistance of the process. The recovery yields of uranium and transuranic elements (TRU) were obtained from a material balance for every unit process composing the whole pyroprocess. The material balances for several elemental groups of interest such as uranium, TRU, rare earth, gaseous fission products, and heat generating elements were calculated on the basis of the knowledge base that is available from domestic and foreign experimental results or technical information presented in open literature. The calculated result of the material balance revealed that uranium and TRU could be recovered at 98.0% and 97.0%, respectively, from a typical PWR spent fuel. Furthermore, the anticipated TRU product was found to emit a non-negligible level of $\gamma$-ray and a significantly higher level of neutrons compared to that of a typical plutonium product obtained from the PUREX process. The results indicate that the product from this conceptual pyroprocessing should be handled in a shielded cell and that this will contribute favorably to retaining proliferation resistance.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2819-2827
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• 2022

The disposal area of a deep geological repository (DGR) for the disposal of spent nuclear fuels (SNFs) is estimated considering the spacing between deposition holes and between disposal tunnels, as determined by a thermal analysis using the decay heat of a reference SNF. Given the relatively large amount of decay heat of the reference SNF, the disposal area of the DGR is found to be overestimated. Therefore, we develop a computer program using MATLAB, termed ACom (Assembly Combination), to combine SNFs when stored in canisters such that the decay heat per canister is evenly distributed. The stability of ACom was checked and the overall distribution of the decay heat per canister was analyzed. Finally, ACom was applied to disposal scenarios suggested in the conceptual design of a DGR for SNFs, and it was confirmed that the decay heat per canister could be evenly distributed and that the maximum decay heat of the canister could be much lower than that of a canister estimated using a reference SNF. ACom can be used to improve the disposal efficiency by reducing the disposal area of a DGR for SNFs by ensuringg a relatively even distribution of decay heat per canister.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2001년도 춘계학술대회 논문집
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• pp.651-654
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• 2001

In the dismantling process of nuclear spent fuels, the spent fuel rod cutting process, followed immediately by the decladding process, performs the cutting the spent fuel rods to a proper length for fast decladding operation. In this paper, we analyzed the chemical compositions, mechanical properties, and physical characteristics for SUS and zircaloy tubes in order to identify the feasibility of cutter-blade type in cutting SUS and zircaloy tubes. It is considered that material, shape and angle, and heat treatment for fabricating the highly durable cutter blade and also it is investigated that the round-shape sustenance of cross-section, amount of debris production, and fire occurrence for measuring the cutting performance on SUS and zircaloy tubes, spent fuel rod cutting device is designed to be operated automatically through the remote control system for use in Hot Cell(radioactive) area and the electro-driven mechanical parts are modularized for easy maintenance. Results from various experiments confirm the efficiency of this device.

• Nuclear Engineering and Technology
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• 제35권3호
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• pp.214-225
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• 2003

A teleoperated manipulator system has been developed for remote handling of the spent fuel bundles. A heavy-duty power manipulator with high reduction ratio joints is used for the slave manipulator in the developed system since the handling tasks of the spent fuel bundles need power. Also, the universal type master manipulator, which has force reflecting capability, is used for precise remote manipulation. The power manipulators so frequently occur the control input saturation that the precise control performances are not achieved due to the windup phenomenon. An advanced bilateral control scheme compensating for the saturation is applied to the teleoperated manipulator system. The validity of the developed system is verified by the grid cutting and fuel transportation tasks from the mockup spent fuel bundle.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 Proceedings of the 4th Korea-China Joint Workshop on Nuclear Waste Management
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• pp.238-244
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• 2004

The relationship between the leaching and gap inventory of spent fuel has been studied. When a specimen of J44H08 spent PWR fuel with 38 GWD/MTU has been leached in the synthetic granitic groundwater in Ar atmosphere, the released fraction of cesium was increased rapidly up to 0.7% at around 500 days and stayed below 0.8% until 3 years. This 0.7% of cesium might be released from the gap in this fuel. The measurement of gap inventory with C15I08 spent PWR fuel, having 35 GWD/MTU and 0.22% of fission gas release, was also determined near 0.6% for the cesium, which is a similar fraction of cesium released from the leaching experiment with J44H08 fuel. Its gap inventories of strontium and iodine were about 0.03 and less than 0.2% respectively. Respective fractions of cesium and strontium in grain boundary of C15I08 were 0.78, 0.09%.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.384.2-384
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.342-342
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• 1999

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.445.1-445
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• 2002