• Applied Chemistry for Engineering
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• v.8 no.1
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• pp.99-107
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• 1997

This study Provides the treatment methods of liquid wastes from the dissolution/purification process of nuclear fuel raw material and the fabrication process of nuclear fuel powder. One of the treatment methods is to process liquid waste from uranium raw material dissolution/purification process. This waste, of the strong acid, can be reused to dissolve the fine ADU particles in filtrate which is ADU waste of pH 8.0 converted from AUC waste after recovery of uranium. To dissolve the fine ADU particles, ADU filtrate was pretreated to pH 4.0 with the dissolution/purification waste, and then mixed with the lime to pH 9.2 and aged for 30 minutes. From this processing, uranium content of the filtrate was decreased to below 3ppm. The waste from fuel powder fabrication is emulsified solution dispersed with fine oil droplets. This emulsion was destroyed effectively by adding and mixing the nitric acid with rapid heating at the same time. After this processing, $Na_2U_2O_7$ compound is produced by addition of NaOH. Optimum condition of this processing was shown at pH 11.5, and uranium content of the filtrate was analyzed to 5ppm. To remove the trace of uranium in the filtrate, lime should be added. Otherwise, 4N nitric acid was used to destroy the emulsion directly, and then lime was added to this waste. Uranium content of the treated filtrate was below 1 ppm. In addition to these wastes, the trace of uranium in filtrate after recovery of uranium from the AUC waste which is produced during PWR power preparation, is treated with NaOH to takeup fluorine(F) in the waste because fluorine is valuable and toxic material. In the finally treated waste, uranium was not detected.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.279-298
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• 2019

This study comprehensively addresses recent progress at KAERI in waste treatment technology to cope with waste produced by pyroprocessing, which is used to effectively manage spent fuel. The goal of pyroprocessing waste treatment is to reduce final waste volume, fabricate durable waste forms suitable for disposal, and ensure safe packaging and storage. KAERI employs grouping of fission products recovered from process streams and immobilizes them in separate waste forms, resulting in product recycling and waste volume minimization. Novel aspects of KAERI approach include high temperature treatment of spent oxide fuel for the fabrication of feed materials for the oxide reduction process, and fission product concentration or separation from LiCl or LiCl-KCl salt streams for salt recycling and higher fission-product loading in the final waste form. Based on laboratory-scale tests, an engineering-scale process test is in progress to obtain information on the performance of scale-up processes at KAERI.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.3
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• pp.457-469
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• 1997

A nuclear spent fuel shipping cask is required by IAEA and domestic regulations to withstand a 9m free drop condition. In this paper, the structural analysis under the 9m side drop condition was performed to understand the dynamic impact behavior and to evaluate the safety of the cask for 7 PWR nuclear spent fuel assemblies. The analysis result was compared with the measured value of the 9m side drop test for the 1/3 scaled-down model and the accuracy of the 3D analysis was confirmed. Analysis in accordance with the diameter of impact limiters for the proto-type cask were performed. Through the analysis, the impact behaviors due to the side drop and the effects dependent on the diameter of impact limiters were grasped. Maximum stress intensities on each part of the cask were respectively calculated by using the stress evaluation program and the structural safety of the cask was finally evaluated in accordance with the regulations.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.77-83
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• 1997

Epoxy resin-type neutron shielding materials, KNS(Kaeri Neutron Shield)-101, KNS-102, and KNS-103 have been fabricated to be used in spent fuel shipping cask. The base material is epoxy resin, and polypropylene, aluminium hydroxide, and boron carbide are added. These shielding materials offer good fluidity at processing, which makes it possible to apply this resin shield to complicated geometric shapes such as shipping cask. The shielding property of these shielding materials for shipping cask for loading 28 PWR spent fuel assemblies has been evaluated. ANISN code is used to evaluate the shielding property of the shipping cask with the thickness of the three neutron shielding materials greater than 10 cm. As a result of analysis, the maximum calculated dose rate at the radial surface of the cask is determined to be $300{\mu}Sv/h$ and the maximum calculated dose rate at 100 cm from the cask is $97{\mu}Sv/h$. These dose rates remain within allowable values specified in related regulations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.233-242
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• 2018

To overcome the low mechanical strength and corrosion behavior of a carbon steel canister at high temperature condition of a deep borehole, SiC ceramics were studied as an alternative material for the disposal canister. In this paper, a design concept for a SiC canister, along with an outer stainless steel container, was proposed, and its manufacturing feasibility was tested by fabricating several 1/3 scale canisters. The proposed canister can contain one PWR assembly. The outer container was also prepared for the string formation of SiC canisters. Thermal conductivity was measured for the SiC canister. The canister had a good thermal conductivity of above $70W{\cdot}m^{-1}{\cdot}K^{-1}$ at $100^{\circ}C$. The structural stability was checked under KURT environment, and it was found that the SiC ceramics did not exhibit any change for the 3 year corrosion test at $70^{\circ}C$. Therefore, it was concluded that SiC ceramics could be a good alternative to carbon steel in application to deep borehole disposal canisters.

• Tunnel and Underground Space
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• v.23 no.2
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• pp.141-149
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• 2013

The thermal analysis is carried out for a geological disposal system developed for the final disposal of a ceramic high-level waste from pyroprocessing of PWR spent fuel. The horizontal disposal tunnel type is considered with the distance of 2 m between the disposal canisters and the tunnel spacing of 25 m. The temperature distributions around the disposal canisters are calculated for the horizontal tunnel based on the conceptual design. The thermal performance analysis is carried out using a FEM program, ABAQUS. The performance analysis shows that the peak temperature in a disposal system outside the disposal canister is lower than $100^{\circ}$, which meets the thermal criterion of the disposal system. According the analysis, the peak temperature for the disposal canister located boundary of the disposal system is lower by $3^{\circ}$ than that for the canister at the central area. This implies the disposal density can be improved by locating more disposal canisters along the boundary.