• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.49-55
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• 2013

To characterize quantitatively the transport of $^{99}Tc$ and the global fallout ($^{137}Cs$, $^{90}Sr$, and $^{239+240}Pu$) for soils in Korea, the transport parameters of a convective-dispersion model, apparent migration velocity, and apparent dispersion coefficient were estimated from the vertical depth profiles of the radionuclides in soils. The vertical profiles of $^{99}Tc$ were measured from a pot experiment for paddy soil that had been sampled from a rice-field around the Gyeongju radioactive waste repository in Korea, and the vertical depth distributions of the global fallout $^{137}Cs$, $^{90}Sr$, and $^{239+240}Pu$ were measured from the soil samples that were taken from local areas in Korea. The front edge of the $^{99}Tc$ profiles reached a depth of about 12 cm in 138 days, indicating a faster movement than the fallout radionuclides. A weak adsorption of $^{99}Tc$ on the soil particles by the formation of Tc(VII) and a high water infiltration velocity seemed to have controlled the migration of $^{99}Tc$. The apparent migration velocity and dispersion coefficient of $^{99}Tc$ for the disturbed paddy soil were 2.88 cm/y and 6.3 $cm^2/y$, respectively. The majority of the global fallout $^{137}Cs$, $^{90}Sr$, and $^{239+240}Pu$ were found in the top 20 cm of the soils even after a transport of about 30 years. The transport parameters for the global fallout radionuclides were 0.01-0.1cm/y ($^{137}Cs$), 0.09-0.13cm/y ($^{90}Sr$), and 0.09-0.18cm/y ($^{239+240}Pu$) for the apparent migration velocity: 0.21-1.09 $cm^2/y$ ($^{137}Cs$), 0.12-0.7$cm^2/y$ ($^{90}Sr$), and 0.09-0.36$cm^2/y$ ($^{239+240}Pu$) for the apparent dispersion coefficient.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.83-91
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• 2018

While using an electrokinetic method to analyze the characteristics of cement solidification of radioactive wastes from decontaminated uranium soil and concrete, the compressive strength, pH, electrical conductivity, irradiation effects, and volume expansion were measured for the solidified cement specimens. The workability of cement solidified from radioactive waste was about 170-190%. After the solidified cement was irradiated, the compressive strength decreased by about 15%, but met the criteria ($34kgf{\cdot}cm^{-2}$) of KORAD (Korea Radioactive Waste Agent). According to the results of SEM-EDS for solidified cement, the aluminum phase was well combined with cement, while the calcium phase was separated from cement. The volume of solidified cement in radioactive wastes was dependent on the waste-to-cement ratio and the amount of water, and increased by about 30% under the conditions used in this study. Therefore, it was concluded that permanent disposal of electrokinetically decontaminated radioactive wastes is appropriate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.123-136
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• 2018

The objective of this study is to propose a new methodology to fabricate a reliable engineering-scale buffer block, which shows homogeneous and uniform distribution in buffer block density, for in-situ experiments. In this study, for the first time in Korea, floating die press and CIP (Cold Isostatic Press) are applied for the manufacture of an engineering-scale bentonite buffer. The optimized condition and field applicability are also evaluated with respect to the method of manufacturing the buffer blocks. It is found that the standard deviation of the densities obtained decreases noticeably and that the average dry density increases slightly. In addition, buffer size is reduced by about 5% at the same time. Through the test production, it is indicated that the stress release phenomenon decreases after the application of the CIP method, which leads to a reduction in crack generation on the surface of the buffer blocks over time. Therefore, it is confirmed that the production of homogeneous buffer blocks on industrial scale is possible using the method suggested in this study, and that the produced blocks also meet the design conditions for dry density of buffer blocks in the AKRS (Advanced Korea Reference Disposal System of HLW).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.239-256
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• 2017

In the process of site selection for a radioactive waste disposal, site characterization must be carried out to obtain input parameters to assess the safety and feasibility of deep geological repository. In this paper, methodologies of site characterization for radioactive waste disposal in Korea were suggested based on foreign cases of site characterization. The IAEA recommends that site characterization for radioactive waste disposal should be performed through stepwise processes, in which the site characterization period is divided into preliminary and detailed stages, in sequence. This methodology was followed by several foreign countries for their geological disposal programs. General properties related to geological environments were obtained at the preliminary site characterization stage; more detailed site characteristics were investigated during the detailed site characterization stage. The results of investigation of geology, hydro-geology, geochemistry, rock mechanics, solute transport and thermal properties at a site have to be combined and constructed in the form of a site descriptive model. Based on this site descriptive model, the site characteristics can be evaluated to assess suitability of site for radioactive waste disposal. According to foreign site characterization cases, 7 or 8 years are expected to be needed for site characterization; however, the time required may increase if the no proper national strategy is provided.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.257-264
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• 2017

For the safe disposal of intermediate level radioactive waste according to the Nuclear Safety and Security Commission's notice and KORAD's management plan for low and intermediate level radioactive waste, the disposal concentration limit was derived based on the IAEA methodology. The evaluation of the derived disposal concentration limit revealed that it is not suitable as a practical limit for intermediate level radioactive waste. This is because the disposal concentration limit according to the IAEA methodology is derived using a single value of radioactive waste density and the disposal facility's volume. The IAEA methodology is suitable for setting the concentration limit for vault type disposal, which consists of a single type of waste, whereas an underground silo type disposal facility is composed of several types of radioactive waste, and thus the IAEA methodology has limitations in determining the disposal concentration limit. It is necessary to develop and apply an improved method to derive the disposal concentration limit for intermediate level radioactive waste by considering the radioactivity of various types of radioactive waste, the corresponding scenario evaluation results, and the regulatory limit.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.265-279
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• 2017

The Gyeongju underground silo type disposal facility, approved for use in December 2014, is in operation for the disposal of low and very low-level radioactive wastes, excluding intermediate-level waste. That is why the existing low-level radioactive waste level has been subdivided and the concentration limit value for intermediate-level waste has been changed in accordance with Nuclear Safety Commission Notice 2014-003. For the safe disposal of intermediate-level wastes, new optimization methodology for calculating the concentration limit of intermediate radioactive level wastes at an underground silo type disposal facility was developed. According to the developed optimization methodology, concentration limits of intermediate-level wastes were derived and the inventory of radioactive nuclides was evaluated. The operation and post closure scenarios were evaluated for the derived radioactive nuclide inventory and the results of all scenarios were confirmed to meet the regulatory limit. However, in case of $^{14}C$, it was confirmed that additional radioactivity limitation through a well scenario was needed in addition to the limit of disposal concentration. It was confirmed that the derived intermediate concentration limit of radioactive waste can be used as the intermediate-level waste concentration limit for the underground disposal facility. For the safe disposal of intermediate-level wastes, KORAD plans to acquire additional data from the radioactive waste generator and manage the cumulative radioactivity of $^{14}C$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.231-237
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• 2017

The present study researched and analyzed decontamination technology for decommissioning a nuclear power plant. The decision-making technique (EXPERT-CHOICE) was used to evaluate and select the optimal decontamination technology. In principle, this evaluation method is generally performed by a group of experts in the relevant field. The results of the weights were calculated by multiplying the weights with regard to each criterion and evaluation score. The evaluation scores were categorized into 3 ranges (high, medium, and low), and each range was weighted for differentiation. The level of the technology analysis was improved by additionally quantifying the weights with regard to each criterion and subdividing criteria into subcriteria. The basic assumption of the evaluation was that the weight values would decided on in an expert survey and assigned to each criterion. The evaluation criteria followed high weight for the 'High' range. Accordingly, H, M, and L were assigned weights of 10:5:1, respectively. This was based on the EXPERT-CHOICE optimal analysis. The minimum and maximum values were excluded, and the average value was used as the evaluation value for each scenario.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.123-133
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• 2010

A statistical evaluation methodology was developed to determine the compliance of candidate waste stream with clearance criteria based upon distribution of radionuclide in a waste stream at a certain confidence level. For the cases where any information on the radionuclide distribution is not available, the relation between arithmetic mean of radioactivity concentration and its acceptable maximum standard deviation was demonstrated by applying widely-known Markov Inequality and One-side Chebyshev Inequality. The relations between arithmetic mean and its acceptable maximum standard deviation were newly derived for normally or lognormally distributed radionuclide in a waste stream, using probability density function, cumulative density function, and other statistical relations. The evaluation methodology was tested for a representative case at 95% of confidence level and 100 Bq/g of clearance level of radioactivity concentration, and then the acceptable range of standard deviation at a given arithmetic mean was quantitatively shown and compared, by varying the type of radionuclide distribution. Furthermore, it was statistically demonstrated that the allowable range of clearance can be expanded, even at the same confidence level, if information on the radionuclide distribution is available.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.215-220
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• 2010

An experiment measuring the concentration ratios of $^{137}Cs$ and $^{85}Sr$ in fish as an index aquatic animal was performed. The species was Chinese minnow (Rhynchocypris Oxycephalus), a Korean native freshwater species. Chinese minnows were reared in acryl aquarium which was 45 cm wide, 85 cm long and 50 cm high. Water in the aquarium was successively purified using filtering devices attached on the floor and the wall. Fish powder in a particulate form was supplied twice a day for feeding. After a radioactive solution was added to make the initial water concentrations approximately $0.02\;{\mu}Ci/l$ and $0.1\;{\mu}Ci/l$ for $^{137}Cs$ and $^{85}Sr$, respectively, the fish and water were sampled 10 times for a month. The concentration ratios were measured to be $0.348lkg^{-1}\sim13.906lkg^{-1}$ for $^{137}Cs$ and $0.474lkg^{-1}\sim13.089lkg^{-1}$ for $^{85}Sr$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.159-166
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• 2010

The block-scale groundwater flow system at Olkiluoto site in Finland was simulated. The heterogeneous and anisotropic hydraulic conductivity field for the domain was constructed from the discrete fracture network, which considered only the fractured zones identified in the deep boreholes installed in the study site. The groundwater flow model was calibrated by adjusting the recharge rate and the transmissivities of the fractured zones to fit the calculated hydraulic heads and into- and out-flow rates in the observation intervals of the boreholes with the observed ones. In the calibrated model, the calculated flow rates at some intervals were not in accordance with the observed ones although the calculated hydraulic heads fit well with the observed ones, which revealed that the number of the conduits for groundwater flow is insufficient in the conceptual model for groundwater flow modeling. Therefore, it was recommended that the potential local conduits such as background fractures should be added to the present conceptual model.