• Journal of the Korean Society of Radiology
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• v.17 no.6
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• pp.947-955
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• 2023

This study was conducted to investigate the awareness of spent nuclear fuel among residents in nuclear power plant areas and use it as basic data for establishing a disposal facility for high-level radioactive waste. 204 questionnaires collected online were analyzed using SPSS Window Ver 28.0. To verify differences between groups, t-test and one-way ANOVA were performed. And correlation analysis was conducted to confirm the relationship between variables. As a result, first, risk perception regarding nuclear-related accidents showed statistically significant differences depending on gender and educational level. The position on the construction of a permanent disposal facility for spent nuclear fuel showed a statistically significant difference depending on gender, education, and age, and the perception of the importance of each evaluation standard for establishing a spent nuclear fuel management plan showed a statistically significant difference depending on education and age. In terms of trust in information-providing institutions, trust in the National Assembly was found to be the lowest. Second, the results of the correlation analysis between variables showed that local residents are aware that an alternative to the current disposal of spent nuclear fuel is needed, and that financial support for the construction of a permanent disposal facility is needed. Therefore, in order to build a high-level radioactive waste disposal site, it is believed that it is necessary to increase trust in the government, collect opinions from local residents, and provide economic support.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.113-119
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• 2011

Safe operation and maintenance of engineered dry storage systems for spent fuel from nuclear power plants basically depends on adequately adopted design requirements. The most important design target of the system are those which provide the necessary assurances that spent fuel can be received, handled, stored and retrieved without undue risk to health and safety of workers and the public. To achieve these objectives, the design of the system incorporates features to remove spent fuel residual heat, to provide for radiation protection, and to maintain containment over the lifespan of the system as specified in the design specifications. The features also provide for all possible anticipated operational occurrences and design basis events in accordance with the design basis as guided by the designated regulations. The general performance requirements of a projected storage system are introduced in this paper. The storage system is designed to store fuel assemblies in associated with designated regulatory requirements. Small increases/decreases in maximum burnup can be adjusted with cooling time. These variations are compensated for by a corresponding small site-specific increase/decrease in the design basis-cooling period, as long as the maximum heat load and radioactivity of loaded fuel assemblies are met. Generic design basis events considered for the storage system are summarized. Shielding and radiological requirements along with mechanical and structural are derived in this study.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.435-435
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.301-302
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• 2005

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.48-55
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• 1989

In this study, simple functional forms which could predict decay heat are referred to and modified in order to analyse more easily long-term behavior of decay heat generated from domestic PWR and CANDU spent fuel. To reduce the difference between the predicted data by functional forms and ORIGEN 2 results and to predict the decay heat under the important parameter(s), sensitivity analysis is performed. By introducing the identified hey parameter, turnup, into the functional forms, the decay heat of spent fuels within a limited rangs of cooling time(3~500 years) becomes predictable for various turnup rates. The predicted decay heat of spent fuels with representative turnup rates such as 33, 37 and 40 GWD/MTU by the functional forms is in so good agreement with ORIGEN 2 results within $\pm$10% difference over the cooling time from 1 to 10$^{5}$ years that the functional forms presented here may be used for engineering purposes such as the thermal design and assessment of the facilities associated with spent fuel management.

• Analytical Science and Technology
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• v.17 no.5
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• pp.381-387
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• 2004

To characterize chemically a spent pressurized water reactor (PWR) fuel, an analytical method for trace amounts of tritium ($^3H$) in it has been established. Considering the effective management of radioactive wastes generated through the whole experimental process and the radiological safety for analysts, a separation condition under which $^{14}C$ and $^3H$ can be sequentially recovered from a single fuel sample was optimized using simulated spent PWR fuel dissolved solutions. $^{14}CO_2$ evolved during dissolution of the spent PWR fuels with nitric acid was trapped in an aliquot of 1.5 M NaOH. $^{129}I_2$ which was volatilized along with $^{14}CO_2$ was removed using a silver nitrate-impregnated silica gel absorbent. $^3H$ remaining in the fuel dissolved solution as $^3H_2O$ was selectively recovered by distillation. Its recovery yield was 97.9% with a relative standard deviation of 0.9% (n=3). $^3H$ in a spent PWR fuel with burnup value of 37,000 MWd/MtU was analyzed, reliability of this analytical method being evaluated by standard addition method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.233-244
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• 2008

A conceptual design study for a pyroprocesing facility, has been carried out in this study, which is available for the recovery of uranium and transuranic elemental group(TRU), that is, reusable as a nuclear fuel especially in a next generation-type reactor. The scale of this facility has been chosen as 20 kg HM/batch, comparatively small engineering size in order to collect scale-up data for the design of a commercial facility as well as to get operational experience. The spent fuel to be handled in this process is as follows : 3.5 % enriched uranium fuel, 35,000MWd/tU and 5-year cooled. The major items considered in the conceptual study are a building lay-out including various hot cells, safety management of the process operation in conjunction with material balance control, radiation safety, inert atmosphere control in shielded hot cells, and criticality control of uranium and TRU products.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.3
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• pp.201-209
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• 2015

Currently, 23 nuclear power plants are in operation at Kori, Uljin, Younggwang and Wolsong site and a reference deep geological disposal system has been developed for the spent fuels generated by them. The reference spent fuel for this disposal system has 4.5wt% of initial enrichment, 55 GWd/MtU of burn-up, and 40 years of cooling time. In this paper, to improve disposal efficiency and economic feasibility, the characteristics of spent fuels from nuclear power plants, such as type and burn-up, were reviewed. A disposal canister concept for shorter length and relatively lower burn-up spent fuels than the reference spent fuels was developed. Based on this canister concept, thermal analyses were carried out and a deep geological disposal concept was proposed. Measures of disposal efficiency such as unit disposal area and disposal density were compared between this disposal system and the reference disposal system. Also, economic feasibility, such as the volume reduction of copper, cast iron, and bentonite, was analyzed and the results of these analyses showed that the disposal system proposed in this paper has an efficiency of at least 20%. These results could be used for establishing spent fuel management policy and designing practical disposal systems for spent fuels.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2104-2125
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• 2021

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three history indices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, this method can directly consider three-dimensional core simulation conditions using history indices of the operating conditions. Therefore, this method reduces the computation time by avoiding a recalculation of the fuel depletion. The module for isotope inventory calculates the number densities using the Lagrange interpolation method and power history correction factors, which are applied to correct the effects of the decay and fission products generated at different power levels. To assess the reliability of the developed code system for back-end cycle analysis, validation study was performed with 58 measured samples of pressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF, HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can provide reasonable results with comparable confidence intervals. As a result, this paper successfully demonstrates that the isotope inventory prediction code system can be used for spent nuclear fuel analysis.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.327-362
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• 2005

The design, construction, and operation of a light water reactor (LWR) are subject to compliance with safety criteria specified for accident situations, such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA). Because reactor fuel is the primary source of radioactivity and heat generation, such a criterion is established on the basis of the characteristics and performance of fuel under the specific accident condition. As such, fuel behavior under accident situations impact many aspects of fuel design and power generation, and in an indirect manner, even spent fuel storage and management. This paper provides a comprehensive review of: the history of the current LOCA criteria, results of LOCA-related investigations on conventional and new classes of fuel, and status of on-going studies on high-burnup fuel under LOCA situations. The objective of the paper is to provide a better understanding of important issues and an insight helpful to establish new LOCA criteria for modem LWR fuels.