• The Journal of Engineering Geology
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• v.33 no.2
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• pp.335-353
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• 2023

Site selection processes for high-level radioactive waste disposal facilities in different countries differ in terms of local geology and degree of public engagement. There seem to be three alternative processes for site selection: (1) selection with community consent after government choice; (2) selection with continuous community engagement after exclusion of unsuitable areas based on existing survey data; or (3) site selection where communities have expressed a willingness to participate. The Yucca Mountain site in Nevada, USA, was selected as the final disposal site by process (1) through six stages, but its development was suspended owing to opposition from the local governor and environmental groups. In Sweden, Switzerland, and Germany, process (2) is used and sites are selected through three stages. Sweden and Switzerland have completed site selection, and Germany is currently engaged in the process. The UK adopted process (3) with six stages, although the process has been suspended owing to poor community participation. In Korea, temporary storage facilities for spent nuclear fuel will reach saturation from 2030, so site selection must be promoted through various laws and systems, with continuous communication with local communities based on transparent and scientifically undertaken procedures.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.105-115
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• 2012

Necessity of demonstration test to evaluate the structural integrity of a basket for accident conditions arose during license approval procedure for the WSPP's dry storage facility named MACSTOR/KN-400. A drop test facility for demonstration was constructed in KAERI site and demonstration tests for basket drop were conducted. As the upper welding region of a loaded basket was collided with a dropped basket during the drop test, the welding in this region was fractured and leakage happened after the drop test. The enhancement of basket design was needed since the existing basket design was not able to satisfy the performance requirement. The directions for design modification were determined and six enhanced designs were derived based on these directions. Structural analyses and specimen tests for each enhanced design were conducted. By evaluating structural analysis results and test results, one among six enhanced designs was decided as a final design for revision. The final design was the one to reduce the height of central post of a basket and to decrease the impact velocity with a dropped basket. Test basket models were fabricated with accordance with the final enhanced design. Additional demonstration test was performed for this test model and all the performance requirements were satisfied.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Journal of Radiation Industry
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• v.18 no.3
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• pp.249-253
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• 2024

For the safe management and disposal of spent fuel, it is essential to accurately determine the quantities of nuclides present within the spent fuel. In this study, a Gamma/Neutron detection system was developed as a part of basic research to measure the burnup profile of spent fuel, and a performance was evaluated using major nuclides. The prototype of the Gamma/Neutron detection system consisted of a CZT sensor and a ³He chamber. For quantitative evaluation, studies were conducted using calibrated ¹³⁷Cs, ¹³⁴Cs, ¹⁵⁴Eu and ²⁵²Cf sources. In the performance evaluation, a field applicability was verified by analyzing the detection characteristics according to the nuclide.

• Journal of Radiation Industry
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• v.18 no.3
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• pp.255-266
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• 2024

The uranium nuclear fuel used in nuclear power generation needs to be replaced with new fuel after a certain period. In South Korea, the spent nuclear fuel generated during this process is temporarily stored within the nuclear power plant site, and there are ongoing issues with the saturation of storage capacity. To address these problems, the South Korea government has established a plan to manage high-level radioactive waste, including provisions for securing interim storage facilities. An interim storage facility is designed to safely store spent nuclear fuel for certain period before its permanent disposal. This study analyzed leading international cases of interim storage facilities that are technically feasible and can reduce the operating period of temporary storage facilities for spent nuclear fuel within nuclear power plant sites. It also presented the technical concepts required for the operation of interim storage facilities for spent fuel from PWR(Pressurized Water Reactor), reflecting the situation in South Korea.

• Journal of the Korea Concrete Institute
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• v.26 no.1
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• pp.71-78
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• 2014

The long term integrity of concrete cask is very important for spent nuclear fuel dry storage system. However, there are serious concerns about early deterioration of concrete cask from creaking and corrosion of reinforcing steel by chloride ion because the cask is usually located in seaside, expecially by combined deterioration such as chloride ion and heat, carbonation. This study is to investigate the relation between temperature and chloride ion diffusion of concrete. Immersion tests using 3.5% NaCl solution that were controlled in four level of temperature, i.e. 20, 40, 65, and $90^{\circ}C$, were conducted for four months. The chloride ion diffusion coefficient of concrete was predicted based on the results of profiles of Cl- ion concentration with the depth direction of concrete specimens using the method of potentiometric titration by $AgNO_3$. Test results indicate that the diffusion coefficient of chloride ion increases remarkably with increasing temperature, and there was a linear relation between the natural logarithm values of the diffusion coefficients and the reciprocal of the temperature from the Arrhenius plots. Activation energy of concrete in this study was about 46.6 (W/C = 40%), 41.7 (W/C = 50%), 30.7 (W/C = 60%) kJ/mol under a temperature of up to $90^{\circ}C$, and concrete with lower water-cement ratio has a tendency towards having higher temperature dependency.

• Journal of Radiation Industry
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• v.17 no.3
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• pp.283-292
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• 2023

Spent nuclear fuel corresponds to high-level radioactive waste that has high decay heat and radioactivity. Accordingly, Spent nuclear fuel withdrawn from the reactor core is primarily stored and managed in a spent nuclear fuel pool in the nuclear power plant to reduce decay heat and radioactivity. In Korea, most nuclear power plant store all spent nuclear fuel in a spent nuclear fuel pool. For wet storage, there are no defense in depth different with reactor core. The study related to spent nuclear fuel pool accident should be carried out to ensure safety. Therefore, it is necessary to analyze previous study cases related to safety of spent nuclear fuel pool and accident cases to build foundational knowledge. The Objective of this study is to analyze study cases of safety assessment and cases for spent nuclear fuel pool accident. For analyzing study cases of safety assessment, possible phenomena when spent nuclear fuel pool accident occurring identified, Subsequently, study cases for safety assessment about each phenomena were investigated, and materials & methods and results for each study are analyzed. For analyzing cases for spent nuclear fuel pool accident, we analyzed accident cases caused by loss of cooling and loss of coolant in spent nuclear fuel pool. Subsequently, causes and change of water level and temperature by each accident case are analyzed. As a result of the analysis on study cases of spent nuclear fuel pool accident, the results of the study conducted by each research institute were vary depending on the computer code, materials & methods of experiment and major assumptions used in the study. As a result of analyzing cases for spent nuclear fuel pool accident, it was found that accident cases for loss of cooling is more than cases for loss of coolant accident. Even though the types of accident in spent nuclear fuel pool were similar, the specific causes were different by each accident case. All the accident cases analyzed did not lead to severe accidents, such as nuclear fuel being exposed to the air. The result of this study will be used as fundamental data for study on spent nuclear fuel pool accident that will be conducted in the future.