• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.637-643
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• 2022

Uranium and plutonium are required to be accounted in spent fuel head-end and major recovery area in pyro-process for safeguards purpose. The possibility of neutron resonance technique, as a nondestructive analysis, was simulated on isotopic fissile analysis for large scale process. Neutron resonance technique has advantage to distinguish uranium from plutonium directly in mixture. Simulation was performed on U235 and Pu239 assay in spent fuel and for scoping examination of assembly type. The resonance energies were determined for U235 and Pu239. The linearity in the neutron transmission was examined for the selected resonance energies. In addition, the limit for detection was examined by changing sample density, thickness and content for actual application. Several factors were proposed for neutron production and the moderated neutron source was simulated for effective and efficient transmission measurement. From the simulation results, neutron resonance technique is promising to analyze U235 and Pu239 for spent fuel assembly. An accurate fissile assay will contribute to an increased safeguards for the pyro-processing system and international credibility on the reuse of fissile materials in the fuel cycle.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.248-255
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• 1983

The thermal analysis on the spent fuel shipping cask for a PWR fuel assembly is performed. Under the normal and fire-accident conditions the temperature distribution through a multilayer cask calculated in compliance with 10 CFR Part 71. A KNU 5&6 spent fuel assembly is assumed to be the decay heat source, which has the maximum discharge turnup of 45, 000MWD/MTU and has been stored in the spent fuel storage pool for 300 days. As a result of thermal analysis, the maximum cladding temperature in case of dry cavity under fire-accident conditions is calculated to be 455$^{\circ}C$. This value is much less than the limiting value specified in 10 CFR Part 50.46. It indicates that no fuel rod cladding rupture could occur under fire-accident conditions. It was also found that no melting of lead would take place in the major shield region.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1457-1462
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• 2017

Unsteady thermal processes in storage containers with spent nuclear fuel were modeled. The daily fluctuations of outer ambient temperatures were taken into account. The modeling approach, which is based on the solving of conjugate and inverse heat transfer problems, was verified by comparison of measured and calculated temperatures in outer channels. The time delays in the reaching of maximal temperatures for each spent fuel assembly were calculated. Results of numerical investigations show that daily fluctuation of outer temperatures does not have a large influence on the maximal temperatures of stored spent fuel, so that fluctuation can be neglected and only daily average temperature should be considered for safety estimation using the "best estimation" approach.

• Journal of the Korean Society for Nondestructive Testing
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• v.22 no.1
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• pp.9-13
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• 2002

A pool image processing measurement method has been developed to improve the examination efficiency and to minimize the errors of dimensional measurements of spent fuel assemblies in pool. Diameter and length measurements of mock-up fuel rods using the image processing system are $-0.24{\pm}0.03mm,\;0.34{\pm}0.06mm$ on the basis of the true value and their maximum errors are within -0.3 and 0.4mm, respectively, According to the result of dimensional measurement of spent fuels in pool, the upper and lower part diameter and mid part diameter of fuel rods of the J44 fuel assembly irradiated for 2 cycles in the Kori-2 nuclear reactor were decreased by about 2.0 and 3.0% in comparison with design values, respectively. The length of fuel rods was elongated by about 0.4%. The change behavior of diameter and length. of fuel rods of the F02 fuel assembly irradiated for 3 cycles in the Kori-1 nuclear reactor showed a trend similar to the results of J44.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1333-1344
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• 2019

An integrated approach of model simplification for high burnup spent nuclear fuel is proposed based on material calibration using optimization. The spent fuel rods are simplified into a beam with a homogenous isotropic material. The proposed approach of model simplification is applied to fuel rods with two kinds of interfacial configurations between the fuel pellets and cladding. The differences among the generated models and the effects of interfacial bonding efficiency are discussed. The strategy of model simplification adopted in this work is to force the simplified beam model of spent fuel rods to possess the same compliance and failure characteristics under critical loads as those that result in the failure of detailed fuel rod models. It is envisioned that the simplified model would enable the assessment of fuel rod failure through an assembly-level analysis, without resorting to a refined model for an individual fuel rod. The effective material properties of the simplified beam model were successfully identified using the integrated optimization process. The feasibility of using the developed simplified beam models in dynamic impact simulations for a horizontal drop condition is examined, and discussions are provided.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.149-158
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• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2288-2297
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• 2023

Current designs of both large reactor units and small modular reactors utilize a nuclear fuel with increasing enrichment. This increasing demand for better nuclear fuel utilization is a challenge for nuclear fuel handling facilities. The operation with higher enriched fuels leads to reduced reserves to legislative and safety criticality limits of spent fuel transport, storage and final disposal facilities. Design changes in these facilities are restricted due to a boron content in steel and aluminum alloys that are limited by rolling, extrusion, welding and other manufacturing processes. One possible solution for spent fuel pools and casks is the burnup credit method that allows decreasing very high safety margins associated with the fresh fuel assumption in spent fuel facilities. This solution can be supplemented or replaced by an alternative solution based on placing the neutron absorber material directly into the fuel assembly, where its efficiency is higher than between fuel assemblies. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the fuel assemblies. The paper summarizes possibilities of fixed neutron absorbers for various nuclear fuel and fuel handling facilities. Moreover, an absorber material was optimized to propose alternative options to boron. Multiple effective absorbers that do not require steel or aluminum alloy compatibility are discussed because fixed absorbers are placed inside zirconium or steel cladding.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.43-44
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.125-126
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• 2017

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.193-204
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• 2023

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO₂ fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.