• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.255-269
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• 2023

Compared to operational wastes, nuclear power plant (NPP) decommissioning wastes are generated in larger quantities within a short time and include diverse types with a wider range of radiation characteristics. Currently used 200 L drums and IP-2 type transport containers are inefficient and restrictive in packaging and transporting decommissioning wastes. Therefore, new packaging and transport containers with greater size, loading weight, and shielding performance have been developed. When transporting radioactive materials, radiological safety should be assessed by reflecting parameters such as the type and quantity of the package, transport route, and transport environment. Thus far, safety evaluations of radioactive waste transport have mainly targeted operational wastes, that have less radioactivity and a smaller amount per transport than decommissioning wastes. Therefore, in this study, the possible radiation effects during the transport from NPP to disposal facilities were evaluated to reflect the characteristics of the newly developed containers and decommissioning wastes. According to the evaluation results, the exposure dose to transport workers, handling workers, and the public was lower than the domestic regulatory limit. In addition, all exposure dose results were confirmed, through sensitivity analysis, to satisfy the evaluation criteria even under circumstances when radioactive materials were released 100% from the container.

• The Journal of Engineering Geology
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• v.27 no.4
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• pp.463-474
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• 2017

Heterogeneous background fractures of granite and sedimentary rocks in Gyeongju LILW (low-intermediate level radioactive waste) facility area have been characterized quantitatively by analyzing fracture parameters (orientation, intensity, and size). Surface geological survey, electrical resistivity survey, and acoustic televiewer log data were used to characterize the heterogeneity of background fractures. Bootstrap method was applied to represent spatial anisotropy of variably oriented background fractures in the study area. As a result, the fracture intensity was correlated to the inverse distance from the faults weighted by nearest fault size and the mean value of electrical resistivity and the average volumetric fracture intensity ($P_{32}$) was estimated as $3.1m^2/m^3$. Size (or equivalent radius) of the background fractures ranged from 1.5 m to 86 m and followed to power-law distribution based on the fractal property of fracture size, using fractures measured on underground silos and identified surface faults.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.253-266
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• 2016

In accordance with the classification system of radioactive waste in Korea, all the disused sealed radioactive sources (DSRSs) fall under the category of EW, VLLW or LILW, and should be managed in compliance with the restrictions for the disposal method. In this study, the management and disposal method are drawn in consideration of half-life of radionuclides contained in the source and A/D value (i.e. the activity A of the source dividing by the D value for the relevant radionuclide, which is used to provide an initial ranking of relative risk for sources) in addition to the domestic classification scheme and disposal method, based on the characteristic analysis and review results of the management practices in IAEA and foreign countries. For all the DSRSs that are being stored (as of March 2015) in the centralized temporary disposal facility for radioisotope wastes, applicability of the derivation result is confirmed through performing the characteristic analysis and case studies for assessing quantity and volume of DSRSs to be managed by each method. However, the methodology derived from this study is not applicable to the following sources; i) DSRSs without information on the radioactivity, ii) DSRSs that are not possible to calculate the specific activity and/or the source-specific A/D value. Accordingly, it is essential to identify the inherent characteristics for each of DSRSs prior to implementation of this management and disposal method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.311-324
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• 2013

To ensure the reliability of computer programs used for the post-closure safety assessment in the Wolsong LILW Center, the results from MASCOT, SAFE-ROCK and GOLDSIM programs are compared with a problem for degradation. Advantages and disadvantages of each computer programs are individually analyzed. Effects on the individual dose are assessed with each computer programs. MASCOT and SAFE-ROCK showed similar results for $^{129}I$ and $^3H$. However, GOLDSIM represented different results for $^{129}I$ and $^3H$. It is analyzed further and compared with the fluxes in each barrier of the disposal system. Througout the benchmarking testing of the computer program, the limitation of computer program can be continuously found out for the mature post-closure safety of Korean radwaste disposal system.

• Journal of Korean Society of Environmental Engineers
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• v.37 no.8
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• pp.465-471
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• 2015

Adsorption experiments for radionuclides such as $^3H$, $^{90}Sr$ and $^{99}Tc$ were conducted using fractured rock collected in unsaturated zone. The released radionuclide through artificial barrier from the near surface repository can be transported by the flow of rainfall or pore water through fractures in unsaturated zone and reach to groundwater flow. Therefore, it is important to investigate transport behavior (retardation) of radionuclides through fractured rock for the safety assessment and long-term performance of repository. Fractured rock samples were collected and characterized by X-ray microtomography (XMT) analysis, which can be used to develop a more robust unsaturated fracture transport model. When fracture-filling materials are exist, distribution coefficient of $^{90}Sr$ is higher than without fracture-filling materials. In this study, batch sorption distribution coefficient ($K_d$) of radionuclide was determined and used to increase our understanding of radionuclide retardtion through fracture-filling materials.