• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.163-170
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• 2023

Since spent nuclear fuel assemblies (SFA) are transported to interim storage or final disposal facility after cooling the decay heat, finite element analysis (FEA) with simplification is widely used to show their integrity against cladding failure to cause dispersal of radioactive material. However, there is a lack of research addressing the comprehensive impact of shape and element simplification on analysis results. In this study, for the optimization of a typical pressurized water reactor SFA, different types of finite element models were generated by changing number of fuel rods, fuel rod element type and assembly length. A series of FEA in use of these different models were conducted under a shock load data obtained from surrogate fuel assembly transportation test. Effects of number of fuel rods, element type and length of assembly were also analyzed, which shows that the element type of fuel rod mainly affected on cladding strain. Finally, an optimal finite element model was determined for other practical application in the future.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2012.05a
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• pp.1019-1020
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• 2012

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2012.10a
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• pp.719-720
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• 2012

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2009.10a
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• pp.413-414
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• 2009

• 제어로봇시스템학회:학술대회논문집
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• 1987.10b
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• pp.613-618
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• 1987

Rod consolidation is a method of increasing spent nuclear fuel storage capacity by disassembling fuel assemblies thus storing the fuel rods in a tighter array. It involves some basic operations which closely resemble to the material handling of a manufacturing process. But all the operations must be controlled remotely in shielded environment from outside due to the highly radioactive nature of the workpiece. In this study the status of the rod consolidation technology in other countries has been surveyed and a feasibility study for the conceptual design of this process have been made.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2001.04a
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• pp.651-654
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• 2001

In the dismantling process of nuclear spent fuels, the spent fuel rod cutting process, followed immediately by the decladding process, performs the cutting the spent fuel rods to a proper length for fast decladding operation. In this paper, we analyzed the chemical compositions, mechanical properties, and physical characteristics for SUS and zircaloy tubes in order to identify the feasibility of cutter-blade type in cutting SUS and zircaloy tubes. It is considered that material, shape and angle, and heat treatment for fabricating the highly durable cutter blade and also it is investigated that the round-shape sustenance of cross-section, amount of debris production, and fire occurrence for measuring the cutting performance on SUS and zircaloy tubes, spent fuel rod cutting device is designed to be operated automatically through the remote control system for use in Hot Cell(radioactive) area and the electro-driven mechanical parts are modularized for easy maintenance. Results from various experiments confirm the efficiency of this device.

• Journal of Korea Foundry Society
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• v.20 no.2
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• pp.80-88
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• 2000

Continuous casting equipment was designed to cast the metallic uranium rods, and a thermal analysis was carried out to calculate the temperature and solidification profiles. Fluid flow and heat transfer analysis model including the effects of phase change was used to simulate the continuous casting process by finite volume method. In the design of continuous casting equipment, the casting speed, pouring temperature and cooling conditions should be considered as significant factors. In this study, the effects of casting speed, pouring temperature, and air gap between the uranium and mold were investigate. The results represented that the temperature and solidification profiles of continuous casting equipment varied with the casting speed, pouring temperature, and air gap.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2412-2420
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• 2024

The International Atomic Energy Agency (IAEA) mandates safeguards to ensure non-proliferation of nuclear materials. Among inspection techniques used to detect partial defects within spent nuclear fuel (SNF), gamma emission tomography (GET) has been reported to be reliable for detection of partial defects on a pin-by-pin level. Conventional GET, however, is limited by low detection efficiency due to the high density of nuclear fuel rods and self-absorption. This paper proposes a new type of GET named Spent Fuel Internal Tomography (SFIT), which can acquire sinograms at the guide tube. The proposed device consists of the housing, shielding, C-shaped collimator, reflector, and gadolinium aluminum gallium garnet (GAGG) scintillator. For accurate attenuation correction, the source-distinguishable range of the SFIT device was determined using MC simulation to the region away from the proposed device to the second layer. For enhanced inspection accuracy, a proposed specific source-discrimination algorithm was applied. With this, the SFIT device successfully distinguished all source locations. The comparison of images of the existing and proposed inspection methods showed that the proposed method, having successfully distinguished all sources, afforded a 150 % inspection accuracy improvement.

• 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.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4731-4742
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• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.