• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.1195-1196
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.434-435
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• 2005

• Journal of the Computational Structural Engineering Institute of Korea
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• v.14 no.4
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• pp.515-523
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• 2001

In this paper results of a linear structural analysis for design and dimensioning of disposal containers for spent pressurized water reactor nuclear fuel and spent Canadian deuterium and uranium reactor nuclear fuel are presented. The container structure studied here is a solid structure with a cast insert and a corrosion resistant outer shell, which is designed for the spent nuclear fuel disposal in a deep repository. An evenly distributed load of hydrostatic pressure from the groundwater and large swelling pressure from the bentonite buffer are applied on the container. Hence, the container must be designed to endure these large pressure loads. In this study, the array type of inner baskets and thicknesses of outer shell and lid/bottom are attempted to be determined through a linear static structural analysis.

• Journal of Radiation Industry
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• v.18 no.2
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• pp.133-140
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• 2024

The Korean 2^nd basic plan for management of high-level radioactive waste presented a plan to manage spent nuclear fuel through dry storage facilities in NPP on-site. For the construction and operation of the facility, it is necessary to develop the monitoring system of the integrity of spent nuclear fuel before operation. NUREG-1536 recommends that the theoretical cask array, typically in the 2×10 array, should account for shadowing effect among the dry storage casks. The objective of this study was to evaluate neutron flux accounting for shadowing effect among dry storage casks. The neutron release rate was evaluated using ORIGEN based on the design basis fuel condition. And the simulation of dry storage casks and evaluation of the shadowing effect were performed using MCNP. Shadowing effect of other dry storage casks was the highest at the center of the dry storage facility of the 2×10 array compared with the outside of the cask. The shadowing effect of neutron flux on the surface among the metal casks was approximately 18% at point 1, 23% at point 2, and 43% at point 3. For the concrete casks, the shadowing effect of neutron flux on the surface was approximately 46% at point 1, 51% at point 2, and 52% at point 3. This means that correction is necessary to monitor the integrity of spent nuclear fuel in each dry storage cask through evaluation of shadowing effect. The results of this study will be used for comparative analysis of neutron measurement data from spent nuclear fuels in dry storage cask. Additionally, the neutron flux evaluation procedure used in this study could be used as the basic data of safety assessment of dry storage cask and development of safety guide.

• Analytical Science and Technology
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• v.14 no.4
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• pp.291-299
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• 2001

A method of separation and purification of Sm for quantitation of Sm isotopes from various fission products in PWR spent nuclear fuels has been studied. Simulated solution containing inactive metal ions(Cs, Ba, Gd, Eu, Sm and Nd) in place of radioactive fission products was prepared. Sm was separated with 0.5 M $HNO_3$/80% MeOH after washing with 1 M $HNO_3$/90% MeOH on AG $1{\times}8$, anion exchange resin. Sm was purified on cation exchange resin, AG $50W{\times}8$, pretreated with 0.2 M alpha-hydroxisobutyric acid(pH 4.5-4.6) to remove Ba causing isobaric effect Sm from PWR spent fuel. As a result of mass spectrometric measurement, eluted Sm portion did not include isobars form other elements such as Gd, Eu, Pm, Nd and BaO. The contents of Sm and its isotopes($^{147}Sm$, $^{148}Sm$, $^{149}Sm$, $^{150}Sm$, $^{151}Sm$, $^{152}Sm$ and $^{154}Sm$) in spent fuel were determined by isotope dilution mass spectrometric method spiking $^{154}Sm$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.151-159
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• 2017

A deep geological disposal at a depth of 500 m in stable host rock is considered to be the safest method with current technologies for disposal of spent fuels classified as high-level radioactive waste. The most important requirement is that the temperature of the bentonite buffer, which is a component of the engineered barrier, should not exceed $100^{\circ}C$. In Korea, the amount of spent fuel generated by nuclear power generation, which accounts for about 30% of the total electricity, is continuously increasing and accumulating. Accordingly, the area required to dispose of it is also increasing. In this study, various duplex disposal concepts were derived for the purpose of improving the disposal efficiency by reducing the disposal area. Based on these concepts, thermal analyses were carried out to confirm whether the critical disposal system requirements were met, and the thermal stability of the disposal system was evaluated by analyzing the results. The results showed that upward 75 m or downward 75 m apart from the reference disposal system location of 500 m depth would qualify for the double layered disposal concept. The results of this study can be applied to the establishment of spent fuel management policy and the design of practical commercial disposal system. Detailed analyses with data of a real disposal site are necessary.

• Analytical Science and Technology
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• v.13 no.6
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• pp.751-758
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• 2000

A study was carried out on the dissolution of spent PWR fuels and performed on the fuels and the separation of iodide for the quantitative analysis using SIMFUEL which has chemical composition of a simulated spent PWR fuel (burn-up; 35,000 MWd/MTU and cooling time; 10 years). To dissolve the SIMFUEL effectively and to minimize the formation of volatile iodine through dissolution process, the optimum ratio of mixed acid ($HNO_3/HCl$ 80: 20 mol%) was established and ozone gas was purged. In the separation step of iodine with $CCl_4$, $NH_2OH{\cdot}HCl$ was used for reducing ${IO_3}^-$ to $I_2$.The optimum acidity of the dissolved solution and the added of $NH_2OH{\cdot}HCl$ were 2.5 M and more than $1.5{\times}10^{-3}mole$, respectively. The recovery of iodide by ion chromatography was $82.8{\pm}4.1%$ and the total yield was corrected by gamma spectrometery using $^{131}I$ as a tracer.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.15 no.3
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• pp.471-480
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• 2002

In this paper, the thermal stress analysis of spent nuclear fuel disposal canister in a deep repository at 500 m underground is carried out for the basic design of the canister. Since the nuclear fuel disposal usually emits much heat, a long term safe repository at a deep bedrock is used. Under this situation, the canister experiences the thermal load due to the heat generation of spent nuclear fuels in the basket. Hence, in this paper the thermal stress analysis is executed using the finite element method. The finite clement code Eot the analysis Is not written directly, but a commercial code, NISA, is used because of the complexity of the structure and the large number of elements required for the analysis. The analysis result shows that even though the thermal stress is added to the stress generated by the hydrostatic underground water pressure and the swelling pressure of the bentonite buffer, the total stress is still smaller than the yield stress of the cast iron. Hence, the canister is still structurally safe when the thermal loads we included in the external loads applied on the canister.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.247-258
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• 2011

Analysis of foreign and domestic patents for pyroprocessing technology of spent nuclear fuels was carried out in this study. The current status of pyroprocessing technology development in such countries as Korea, USA, Japan and EU was analyzed by classifying the patents for 1975 through 2009 according to registration country, assignee, calendar year and technology area. The major assignees' activity indices were compared in order to find out whether there is any concentrated area of technical details. Technology competitiveness of the countries was also investigated from the information of patent citation number and family size. Furthermore, some essential unit technologies required for the commercialization of pyroprocessing were derived and examined in the aspect of the state of art as well as the trend of technology development.

• Analytical Science and Technology
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• v.16 no.6
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• pp.431-437
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• 2003

Metalization yield of uranium oxide to uranium metal from lithium reduction process of spent pressurized water reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metalization yield was measured. Metalization yield of the solid part was 90.7~95.9 wt%, and the powder being 77.8~71.5 wt% individually. Oxidation behaviour of the quartemary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At $600{\sim}700^{\circ}C$, weight increments of alloy of Mo, Ru, Rh and Pd was 0.40~0.55 wt%. Phase change on the surface of the alloy was started at $750^{\circ}C$. In particular, Mo was rapidly oxidized and then the alloy lost 0.76~25.22 wt% in weight.