• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.18 no.spc
• /
• pp.1-19
• /
• 2020

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

• Nuclear Engineering and Technology
• /
• v.24 no.2
• /
• pp.178-182
• /
• 1992

Spent PWR fuel rods have been scanned axially and sectionally to measure the relative gamma-ray intensity of Cs-137 and then bumups of the scanned rods determined by measuring Nd-148 which has been chemically separated. From these experimental results, a linear relation(LR) between the gamma-ray intensity of Cs-137 and the bumup in the range of 10∼35 GWD/MTU was obtained. In order to validate the LR, the Cs-137 gamma-ray intensity of unknown sample was nondestructively measured and the bumup obtained by the LR was compared with that of the Nd-148 method. It is revealed that the results from both methods are in good agreement, and thus it seems to be possible to estimate the bumup of spent PWR fuel rod by measuring nondestructively gamma-ray of fission product Cs-137.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.19 no.4
• /
• pp.491-501
• /
• 2021

This paper aims to evaluate the mechanical integrity for Spent Nuclear Fuel (SNF) cladding under lateral loads during transportation. The evaluation process requires a conservative consideration of the degradation conditions of SNF cladding, especially the hydride effect, which reduces the ductility of the cladding. The dynamic forces occurring during the drop event are pinch force, axial force and bending moment. Among those forces, axial force and bending moment can induce transverse tearing of cladding. Our assessment of 14 × 14 PWR SNF was performed using finite element analysis considering SNF characteristics. We also considered the probabilistic procedures with a Monte Carlo method and a reliability evaluation. The evaluation results revealed that there was no probability of damage under normal conditions, and that under accident conditions the probability was small for transverse failure mode.

• Nuclear Engineering and Technology
• /
• v.42 no.1
• /
• pp.79-88
• /
• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.6 no.2
• /
• pp.147-154
• /
• 2008

It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.

• Nuclear Engineering and Technology
• /
• v.33 no.2
• /
• pp.231-240
• /
• 2001

When it is assumed that PWR, CANDU and DUPIC spent fuels are disposed of in deep geological repository, consequent annual individual doses are calculated, and it is shown that doses meet the regulatory limit. From these results, the hazardous radionuclides applicable to partitioning and transmutation are selected. These selected radionuclides such as Tc-99, Ⅰ-129, Cs-135 and Np-237 are then reviewed in terms of partitioning and transmutation. Separation of I-129, Np-237 and Tc-99 from spent fuels is considered desirable, and transmutation of these radionuclides results in remarkable hazard reduction. However, it is concluded that separation and transmutation of Cs-135 may be ineffective although it is classified into a hazardous radionuclide.

• Proceedings of the Computational Structural Engineering Institute Conference
• /
• 2001.04a
• /
• pp.259-266
• /
• 2001

This paper presents the results of a structural analysis to determine design variables such as the inner basket array type, and thicknesses of the outer shell and the lid and bottom of a spent nuclear fuel disposal canister. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock, entailing an evenly distributed load of hydrostatic pressure from the groundwater and large swelling pressure from the bentonite buffer. Hence, the canister must be designed to withstand these large pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables, the array type of inner baskets and thicknesses of outer shell and lid and bottom are attempted to be determined through a linear static structural analysis. Canister types studied here are one for the pressurized water reactor (PWR) fuel and another for the Canadian deuterium and uranium reactor (CANDU) fuel.

• Analytical Science and Technology
• /
• v.17 no.5
• /
• pp.438-442
• /
• 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)
• /
• v.22 no.3
• /
• pp.347-362
• /
• 2024

In Korea, two types of spent nuclear fuels (SNFs) are generated, pressurized light water reactor type (PWR) and pressurized heavy water reactor type (PHWR; CANDU), that differ greatly in size, decay heat, and radioactive characteristics. Technology development for the disposal of SNFs has mainly focused on PWR SNFs that are large in size and have extremely high decay heat and radioactivity. However, CANDU SNFs should be considered differently from PWR SNFs in deep geological disposal systems because their characteristics significantly differ from those of PWR SNFs in terms of their dimensions, number of SNF bundles, and handling systems in nuclear power plant sites. In this paper, after reviewing the status of the CANDU SNF disposal concept by Canada and Korea, concepts related to the direct geological disposal of CANDU SNFs were described, and two concepts were proposed based on the results of the development. The engineered barrier systems developed using these two concepts were comparatively analyzed in terms of disposal safety, disposal efficiency, and technical maturity. Based on the results of the comparative analyses, a vertical-type emplacement disposal concept was determined as a reference concept for the deep geological disposal of CANDU SNFs.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 2003.06a
• /
• pp.1154-1157
• /
• 2003

The spent fuel slitting device is an equipment developed for the separation of the pellet and hull from the cutting fuel rod with length of 250 mm, and in order to feed UO$_2$ pellet. We have analyzed on the existing technologies for designing and producing of the slitting device in the first year(2001), based on these results, designed and produced the rod slitting device. It has effectively separated the pellet from the hull, but demanded the supplement separation work because of the mixing with pellet and hull in the vessel, and required the condition for the reducing time of the process. In the second year(2002), we have reduced the work time, performed the test and capacity evaluation with the improving device, based these results, and ensured the data demanded for designing of the spent fuel rod slitting device. We have compared with the DUPIC(Direct use of spent PWR fuel in CAND reactors) process, and developed the device for the purpose of reducing over 40 % in comparition with the DUPIC operation time(5 minutes). Based on these results, it will is effectively applied to available data for designing and producing of the hot test facility.