• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.165-169
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• 1988

As a part of tandem fuel cycle feasibility study, the residual U and Pu nuclide contents of PWR spent fuels are computed using ORICEN2 code for each Korea Nuclear Unit and batch to investigate the potential of utilizing them as CANDU fuels. The annual and accumulated discharged amounts of U and Pu nuclides are computed for the PWRs from KNU 1 through KNU 10. The results of computation show that the spent fuels having 0.7-0.8 w/o U-235 are dominant and considerable amounts of fissile Pu are produced. The enrichment of U-235 is less than the expected 0.8-0.9 w/o U-235 since the burnups offered by KEPCO are higher than those of other PWRs.

• Journal of Energy Engineering
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• v.20 no.3
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• pp.236-241
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• 2011

In this study, considering a degree of proliferation resistance and sustainability, development status of perspective recycling systems for spent nuclear fuels (SNF) is comprehensively reviewed on the basis of the urgent needs of sustainable management measures for high level radioactive wastes such as spent nuclear fuels (SNF).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.303-313
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• 2023

Spent fuels (SFs) are stored in a storage pool after discharge from nuclear power plants. They can be transferred to for the further processes such as dry storage sites, processing plants, or disposal sites. One of important measures of SF is the burnup. Since the radioactivity of SF is strongly dependent on its burnup, the burnup of SF should be well estimated for the safe management, storage, and final disposal. Published papers about the methodology for the burnup estimation from the known activities of important radioactive sources are somewhat rare. In this study, we analyzed the dependency of the burnup on the important radiation source activities using ORIGEN-ARP, and suggested simple correlations that relate the burnup and the important source activities directly. A burnup estimation equation is suggested for PWR fuels relating burnup with total neutron source intensity (TNSI), initial enrichment, and cooling time. And three burnup estimation equations for major gamma sources, ¹³⁷Cs, ¹³⁴Cs, and ¹⁵⁴Eu are also suggested.

• Journal of Radiation Protection and Research
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• v.21 no.2
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• pp.117-124
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• 1996

Source intensities in terms of the exposure rates at 1m from the fresh and spent DUPIC fuels, made from standard and extended turnup PWR fuels, were analyzed. Two cases were studied based on the degree of elimination of removable elements. Homogeneous mixture model was applied to get the exposure rate. The exposure rate turned out to be very high and sensitive to Cs elimination during the dry process. About 90% of exposure can be reduced in the case of fresh DUPIC fuel made from 10-year cooled spent PWR fuels if Cs is fully removed during the dry process. The main radiation source in spent fuels is Cs-137. The dry storage of spent DUPIC fuel may need a longer wet storage period and require a further review.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.2
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• pp.255-266
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• 2021

Solid-state mechanochemical reduction combined with subsequent melting consolidation was suggested as a technical option for the oxide reduction in pyroprocessing. Ni ingot was produced from NiO as a starting material through this technique while Li metal was used as a reducing agent. To determine the technical feasibility of this approach for pyroprocessing, which handles spent nuclear fuels, thermodynamic calculations of the phase stabilities of various metal oxides of U and other fission elements were made when several alkaline and alkali-earth metals were used as reducing agents. This technique is expected to be beneficial, not only for oxide reduction but also for other unit processes involved in pyroprocessing.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.21-36
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• 2020

With respect to spent nuclear fuels, disposal containers and bentonite buffer blocks in deep geological disposal systems are the primary engineered barrier elements that are required to isolate radioactive toxicity for a long period of time and delay the leakage of radio nuclides such that they do not affect human and natural environments. Therefore, the thermal stability of the bentonite buffer and structural integrity of the disposal container are essential factors for maintaining the safety of a deep geological disposal system. The most important requirement in the design of such a system involves ensuring that the temperature of the buffer does not exceed 100℃ because of the decay heat emitted from high-level wastes loaded in the disposal container. In addition, the disposal containers should maintain structural integrity under loads, such as hydraulic pressure, at an underground depth of 500 m and swelling pressure of the bentonite buffer. In this study, we analyzed the thermal stability and structural integrity in a deep geological disposal environment of the improved deep geological disposal systems for domestic light-water and heavy-water reactor types of spent nuclear fuels, which were considered to be subject to direct disposal. The results of the thermal stability and structural integrity assessments indicated that the improved disposal systems for each type of spent nuclear fuel satisfied the temperature limit requirement (< 100℃) of the disposal system, and the disposal containers were observed to maintain their integrity with a safety ratio of 2.0 or higher in the environment of deep disposal.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2005.06a
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• pp.1715-1718
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• 2005

During demonstrations of a process conditioning spent nuclear fuels, it may be necessary to transport and handle Spent fuel road cuts from Post Irradiation Examination facility to Slitting device in The hot cell. It may be not easy to transport spent fuel rod cuts because rod cuts are high radioactive materials. For this purpose, we have developed a capsule for transporting and handling high radioactive materials. We have analyzed conditions of a hot cell and requirements of the device, designed and manufactured The prototype of the device, and done some performance tests. From the tests, it has been shown that transportation and handling without scattering nuclear material was smooth but the weight of capsule was heavy. These result will be reflected to a design of the improved transportation and handling device which will be used during demonstrations.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.596-607
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• 2015

In this paper, we describe a computer program, iBEST (inverse Burnup ESTimator), that we developed to accurately estimate the burnup histories of spent nuclear fuels based on sample measurement data. The burnup history parameters include initial uranium enrichment, burnup, cooling time after discharge from reactor, and reactor type. The program uses algebraic equations derived using the simplified burnup chains of major actinides for initial estimations of burnup and uranium enrichment, and it uses the ORIGEN-S code to correct its initial estimations for improved accuracy. In addition, we newly developed a stable bisection method coupled with ORIGEN-S to correct burnup and enrichment values and implemented it in iBEST in order to fully take advantage of the new capabilities of ORIGEN-S for improving accuracy. The iBEST program was tested using several problems for verification and well-known realistic problems with measurement data from spent fuel samples from the Mihama-3 reactor for validation. The test results show that iBEST accurately estimates the burnup history parameters for the test problems and gives an acceptable level of accuracy for the realistic Mihama-3 problems.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2007.11a
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• pp.421-422
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• 2007

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2008.06a
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• pp.785-786
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• 2008