• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.219-228
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• 2012

A long-term R&D program for HLW disposal technology development was launched in 1997 in Korea and Korea Reference disposal System(KRS) for spent fuels had been developed. After then, a recycling process for PWR spent fuels to get the reusable material such as uranium or TRU and to reduce the volume of radioactive waste, called Pyro-process, is being developed. This Pyro-process produces several kinds of wastes including metal waste and ceramic waste. In this study, the characteristics of the waste from Pyro-process and the concepts of a disposal container for the wastes were described. Based on these concepts, thermal analyses were carried out to determine a layout of the disposal area of the ceramic wastes which was classified as a high level waste and to develop the disposal system called A-KRS. The location of the final repository for A-KRS is not determined yet, thus to review the potential repository domains, the possible layout in the geological characteristics of KURT facility site was proposed. These results will be used in developing a repository system design and in performing the safety assessment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.283-295
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• 2007

The technical feasibility of a pyroprocessing of PWR spent fuels to recover nuclear fuel materials, uranium and transuranic elements group(TRU), was examined in this study. Also its applicability as a new fuel cycle technology in terms of non-proliferation was investigated. First, various unit processes were combined to a pyroprocess. Then the flow aspects of such materials of issue as uranium, transuraniums, rare earth, noble metals and heat generating elements were examined on the flowsheet, which was obtained by the assumptions on the basis of various experimental results in this work or separation data collected from literatures. Consequently, the calculated results of the material balance for the whole process showed that uranium and TRU could be recovered as products by 98.0 % and 97.0 %, respectively, from a PWR spent fuel while removing the other elemental groups into radioactive wastes. On the one hand, the TRU product was found to emit a considerable amount of ${\gamma}$-ray as well as neutrons favorably contributing to the strategy of proliferation resistance.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.217-224
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• 2008

In deep geological disposal system, the integrity of a disposal canister having spent fuels is very important factor to assure the safety of the repository system. This disposal canister is one element of the engineered barriers to isolate and to delay the radioactivity release from human beings and the environment for a long time so that the toxicity does not affect the environment. The main requirement in designing the deep geological disposal system is to keep the buffer temperature below 100$^{\circ}C$ by the decay heat from the spent fuels in the canister in order to maintain the integrity of the buffer material. Also, the disposal canister can endure the hydraulic pressure in the depth of 500 m and the swelling pressure of the bentonite as a buffer. In this study, new concept of the disposal canister for the CANDU spent fuels which were considered to be disposed without any treatment was developed and the thermal stability and the structural integrity of the canister were analysed. The result of the thermal analysis showed that the temperature of the buffer was 88.9$^{\circ}C$ when 37 years have passed after emplacement of the canister and the spacings of the disposal tunnel and the deposition holes were 40 m and 3 m, respectively. In the case of structural analysis, the result showed that the safety factors of the normal and the extreme environment were 2.9 and 1.33, respectively. So, these results reveal that the canister meets the thermal and the structural requirements in the deep geological disposal system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.19-32
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• 2010

Electrolytic reduction technology is essential for the purpose of adopting pyroprocessing into spent oxide fuel as an alternative option in a back-end fuel cycle. Spent fuel consists of various metal oxides, and each metal oxide releases an oxygen element depending on its chemical characteristic during the electrolytic reduction process. In the present work, an electrolytic reduction behavior was estimated for voloxidized spent fuel based on the assumption that each metal-oxygen system is independent and behaves as an ideal solid solution. The electrolytic reduction was considered as a combination of a Li recovery and chemical reactions between the metal oxides such as uranium oxide and the produced Li metal. The calculated result revealed that most of the metal oxides were reduced by the process. It was evaluated that a reduced fraction of lanthanide oxides increased with a decreasing $Li_2O$ concentration. However, most of the lanthanides were expected to be stable in their oxide forms. In addition, a semi-empirical model for describing $U_3O_8$ electrolytic reduction behavior was proposed by considering Li diffusion and a chemical reaction between $U_3O_8$ and Li. Experimental data was used to determine model parameters and, then, the model was applied to calculate the reduction yield with time and to estimate the required time for a 99.9% reduction.

• Journal of the Korean Society of Radiology
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• v.15 no.1
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• pp.37-43
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• 2021

Republic of Korea has many nuclear facilities in the country, and Democratic People's Republic of Korea(North Korea) locates in the surrounding country. Therefore, it is necessary to construct the target facility's nuclear forensic data in a preemptive response to the changing international situation. For this reason, this study suggests "signature" materials used to understand the origins and sources of nuclear and other radioactive materials, taking into account domestic nuclear facilities and the nuclear fuel cycle. In domestic, pressurized light water reactors and pressurized heavy water reactors are in operation, and enriched and natural uranium are used as fuels. In the front-end fuel cycle, the signature materials can be nature uranium and UF6 in the uranium enrichment process. The domestic back-end fuel cycle adopts a non-circulating cycle excluding the reprocessing process, and the primary signature material is spent nuclear fuel. According to IAEA recommendation, the importance of these materials as the signature and characteristic contents are suggested in this study. To prove the integrity of nuclear material and build a national nuclear forensics library, it is necessary to grasp the signature material and acquire the characteristic data considering the domestic nuclear facilities and the nuclear fuel cycle.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.339-346
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• 2019

Based on the $8^{th}$ Basic Plan for Electric Power Demand and Supply, an estimation has been made for inventories and characteristics of spent fuel (SF) to be generated from existing and planned nuclear power plants. The characteristics under consideration in this study are dimensions, fuel array, $^{235}U$ enrichment, discharge burnup, and cooling time for each fuel assembly. These are essentially needed for designing a disposal facility for SFs. It appears that the anticipated quantity by the end of 2082 is about 62,500 assemblies for PWR SFs. The inventories of Westinghouse-type and Korean-type SFs were revealed to be 60% and 40%, respectively as of the end of 2018. The proportion of SFs with initial $^{235}U$ enrichment below 4.5 weight percent (wt%) was shown to be approximately 90% in total as of the end of 2018. As of 2077, more than 97% of SFs generated from Westinghouse-type nuclear reactors were shown to have cooling time of over 50 years. As of 2125, more than 98% of SFs generated from Korean-type nuclear reactors were shown to have cooling time of over 45 years. Based on these results, for the efficient design of a disposal system, it is reasonable to adopt two types of reference spent fuel. SF of KSFA with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 50 years was determined as reference fuel for Westinghouse-type SFs; SF of PLUS7 with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 45 years was determined as reference fuel for Korean-type SFs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.245-251
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• 2014

Pyroprocess for treating spent nuclear fuels has been developed based on electrochemical principles. Process simulation is one of the important methods for process development and experimental data analysis and it is also a necessary approach for pyroprocessing. To date, process simulation of pyroprocessing has been focused on electrorefining and there have been not so many investigations on electrolytic reduction. Electrolytic reduction, unlike electrorefining, includes specific features of gas evolution and porous electrode and, thus, different equations should be considered for developing a model for the process. This study summarized required concepts and equations for electrolytic reduction model development from thermodynamic, mass transport, and reaction kinetics theories which are necessitated for analyzing an electrochemical cell. An electrolytic reduction cell was divided and equations for each section were listed and, then, boundary conditions for connecting the sections were indicated. It is expected that those equations would be used as a basis to develop a simulation model for the future and applied to determine parameters associated with experimental data.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.171-188
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• 2012

A GoldSim template program for a safety assessment of a hybrid-typed repository system, called "A-KRS," in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been deterministically assessed with 5 various normal and abnormal scenarios associated with nuclide release and transport in and around the repository. Dose exposure rates to the farming exposure group have been evaluated in accordance with all the scenarios and then compared among other.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.303-309
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• 2010

For the pyrochemical process of spent nuclear fuels, recovery of LiCl-KCl eutectic salts is needed to reduce radioactive waste volume and to recycle resource materials. This paper is about recovery of residual LiCl-KCl eutectic salts in radioactive rare earth precipitates (rare earth oxychlorides or oxides) by using a vacuum distillation process. In the vacuum distillation test apparatus, the salts in the rare earth precipitates were vaporized and were separated effectively. The separated salts were deposited in three positions of the vacuum distillation test apparatus or were collected in the filter and it is difficult to recover them. To resolve the problem, a vacuum distillation and condensation system, which is subjected to the force of a temperature gradient at a reduced pressure, was developed. In a preliminary test of the vacuum distillation/condensation recovery system, it was confirmed that it was possible to condense the vaporized salts only in the salt collector and to recover the condensed salts from the salt collector easily.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.263-272
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• 2012

A GoldSim template program for a safety assessment of a hybrid-typed repository system, called A-KRS, in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been probabilistically assessed with 9 selected input parameters, each of which has its own statistical distribution for a normal release and transport scenario associated with nuclide release and transport in and around the repository. Probabilistic dose exposure rates to the farming exposure group have been evaluated. A sensitivity of 9 selected parameters to the result has also been investigated to see which parameter is more sensitive and important to the exposure rates.