• Journal of Institute of Convergence Technology
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• v.2 no.1
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• pp.7-12
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• 2012

Spent fuels are discharged from nuclear reactors as a result of power generations. The spent fuels would be considered as a useful resources because the main constituent is uranium and some other actinides are included in them. In order to utilize the resources chemical processes should be developed to treat the spent fuels and obtain uranium and other actinides to be fueled in a fast reactor. The technologies are categorized into wet and dry processes. In this study, the current status of such technologies is summarized to give a insight and a deep understanding on nuclear fuel cycles.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2006.05a
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• pp.369-370
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• 2006

During demonstrations of a process conditioning spent nuclear fuels, it is necessary to transport and handle Spent fuel road cuts from Post Irradiation Examination facility to Slitting device in The hot cell. the spent fuel pellets which are highly radioactive materials are separated with its clad and are fed into the next conditioning process. For this, a spent fuel rod, 3.5 m long, is cut by 25 cm long which is suitable length for the decladding process. These rod-cuts are packed into the capsule and are moved to the ACPF(Advanced spent nuclear fuel Conditioning Process Facility). In the ACPF, Once the capsule is unloaded in the ACPF, Capsule is taken out one-by-one and installed on the decladding device. In these processes, the crushed spent fuel pellet can be scattered inside the facilities and thus it contaminate the hot cell. In this paper, we developed the specially designed transportation and storage device for spent nuclear fuel capsules.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.451-463
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• 2023

With South Korea increasingly focusing on nuclear energy, the management of spent nuclear fuel has attracted considerable attention in South Korea. This study established a novel procedure for selecting safety-relevant radionuclides for long-term safety assessments of a deep geological repository in South Korea. Statistical evaluations were performed to identify the design basis reference spent nuclear fuels and evaluate the source term for up to one million years. Safety-relevant radionuclides were determined based on the half-life criteria, the projected activities for the design basis reference spent nuclear fuel, and the annual limit of ingestion set by the Nuclear Safety and Security Commission Notification No. 2019-10 without considering their chemical and hydrogeological properties. The proposed process was used to select 56 radionuclides, comprising 27 fission and activation products and 29 actinide nuclides. This study explains first the determination of the design basis reference spent nuclear fuels, followed by a comprehensive discussion on the selection criteria and methodology for safety-relevant radionuclides.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2008.05a
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• pp.488-490
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• 2008

This study aims to design the high-capacity vol-oxidizer using simulated fuels instead of spent nuclear fuels. Simulated fuels are fabricated by blending tungsten powder with silicon carbide powder, and thereafter, paraffin coating covers simulated fuels to increase their strength. An oxidation experiment using simulated fuels have been carried out in order to analyze oxidation characteristics similar to spent fuels. After oxidation, simulated fuels were almost oxidized to be powders. Increased volume of simulated fuels approached to spent fuels. These results can be utilized as important informations for designing a high-capacity vol-oxidizer.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.84-92
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• 2023

As temporary storage facilities for spent nuclear fuels in domestic nuclear power plants are expected to be saturated, external intermediate storage facilities would be required in the future. Spent nuclear fuels are stored in metal canisters and then placed in a dry environment within concrete or metal casing for operation. In the United States, the dry storage method for spent nuclear fuels has been operated for an extended period. Based on the corrosion experiences of dry storage canisters in chloride environments, numerous studies have been conducted to reduce corrosion in welds. With the construction of intermediate storage facilities in Korea for spent nuclear fuels expected near coastal areas adjacent to nuclear power plants, there is a need for research on the corrosion occurrence of welds and mitigation methods for canisters in chloride environments. In this paper, we measured and compared the residual stresses in the Heat-Affected Zones (HAZ) after electron beam welding (EBW) and gas tungsten arc welding (GTAW) processes for candidate materials such as 304L, 316L, and duplex stainless steel(DSS). We investigated the possibility of microstructure control through the application of surface modification processes using friction stir processing (FSP). Corrosion tests on each welded specimen revealed a higher corrosion rate in EBW welds compared to GTAW. Furthermore, it was confirmed that corrosion resistance improved due to phase refinement and redistribution of precipitates when FSP was applied.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.3
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• pp.201-209
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• 2015

Currently, 23 nuclear power plants are in operation at Kori, Uljin, Younggwang and Wolsong site and a reference deep geological disposal system has been developed for the spent fuels generated by them. The reference spent fuel for this disposal system has 4.5wt% of initial enrichment, 55 GWd/MtU of burn-up, and 40 years of cooling time. In this paper, to improve disposal efficiency and economic feasibility, the characteristics of spent fuels from nuclear power plants, such as type and burn-up, were reviewed. A disposal canister concept for shorter length and relatively lower burn-up spent fuels than the reference spent fuels was developed. Based on this canister concept, thermal analyses were carried out and a deep geological disposal concept was proposed. Measures of disposal efficiency such as unit disposal area and disposal density were compared between this disposal system and the reference disposal system. Also, economic feasibility, such as the volume reduction of copper, cast iron, and bentonite, was analyzed and the results of these analyses showed that the disposal system proposed in this paper has an efficiency of at least 20%. These results could be used for establishing spent fuel management policy and designing practical disposal systems for spent fuels.

• Nuclear Engineering and Technology
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• v.43 no.5
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• pp.413-420
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• 2011

In Korea, although the concept of dry storage system for PWR spent fuels first emerged in the early 1990s, wet storage inside nuclear reactor buildings remains the dominant storage paradigm. Furthermore, as the amount of discharged fuel from nuclear power plants increases, nuclear power plants are confronted with the problem of meeting storage capacity demand. Various measures have been taken to resolve this problem. Dry storage systems along with transportation of spent fuel either on-site or off-site are regarded as the most feasible measure. In order to develop dry storage and transportation system safety analyses, development of design techniques, full scale performance tests, and research on key material degradation should be conducted. This paper deals with two topics, structural analysis methodology to assess cumulative damage to transportation packages and the effects of an aircraft engine crash on a dual purpose cask. These newly emerging issues are selected from among the many technical issues related to the development of transportation and storage systems of spent fuels. In the design process, appropriate analytical methods, procedures, and tools are used in conjunction with a suitably selected test procedure and assumptions such as jet engine simulation for postulated design events and a beyond design basis accident.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.287-297
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• 2014

For several decades, many countries operating nuclear power plants have been studying the various disposal alternatives to dispose of the spent nuclear fuel or high-level radioactive waste safely. In this paper, as a direct disposal of spent nuclear fuels for deep geological disposal concept, the rod consolidation from spent fuel assembly for the disposal efficiency was considered and analyzed. To do this, a concept of spent fuel rod consolidation was described and the related concepts of disposal canister and disposal system were reviewed. With these concepts, several thermal analyses were carried out to determine whether the most important requirement of the temperature limit for a buffer material was satisfiedin designing an engineered barrier of a deep geological disposal system. Based on the results of thermal analyses, the deposition hole distance, disposal tunnel spacing and heat release area of a disposal canister were reviewed. And the unit disposal areas for each case were calculated and the disposal efficiencies were evaluated. This evaluation showed that the rod consolidation of spent nuclear fuel had no advantages in terms of disposal efficiency. In addition, the cooling time of spent nuclear fuels from nuclear power plant were reviewed. It showed that the disposal efficiency for the consolidated spent fuel rods could be improved in the case that cooling time was 70 years or more. But, the integrity of fuels and other conditions due to the longer term storage before disposal should be analyzed.

• Analytical Science and Technology
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• v.17 no.5
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• pp.438-442
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• 2004

To determine the contents of $^{99}Tc$ in the spent PWR (pressurized water reactor) nuclear fuels by ICP-MS (inductively coupled plasma-mass spectrometry), a technetium separation method using an extraction chromatographic resin (TEVA Spec resin) has been established. $^{99}Tc$ was separated from a spent PWR nuclear fuel solution by this separation procedure and its concentration was determined by ICP-MS. The result agrees well with the value calculated by the program ORIGEN 2 and also the value measured by AG MP-1 resin/ICP-MS method described in our previous paper. It can be concluded that the present separation procedure is superior to the AG MP-1 resin procedure with respect to the time required for technetium separation as well as the efficiency of decontamination from other radioactive nuclides.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.503-516
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• 2023

A transfer cask serves as the container for transporting and handling canisters loaded with spent nuclear fuels from light water reactors. This study focuses on a cylindrical transfer cask, standing at 5,300 mm with an external diameter of 2,170 mm, featuring impact limiters on the top and bottom sides. The base of the cask body has an openable/closable lid for loading canisters with storage modules. The transfer cask houses a canister containing spent nuclear fuels from lightweight reactors, serving as the confinement boundary while the cask itself lacks the confinement structure. The objective of this study was to conduct a structural analysis evaluation of the transfer cask, currently under development in Korea, ensuring its safety. This evaluation encompasses analyses of loads under normal, off-normal, and accident conditions, adhering to NUREG-2215. Structural integrity was assessed by comparing combined results for each load against stress limits. The results confirm that the transfer cask meets stress limits across normal, off-normal, and accident conditions, establishing its structural safety.