• Journal of the Korean institute of surface engineering
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• v.28 no.5
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• pp.259-266
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• 1995

In order to manufacture an attractive waste glass for the permanent and secure disposal of high-level radioactive waste, the complex composition of the simulated nuclear waste glass PNL-7668 was simplified to a composition of sodium borosilicate glass. The substitutions of $Fe_2O_3$ and $Al_2O_3$ were added to examine on the leaching characteristics of simulated nuclear waste glass with variable composition. The leach tests for these glasses were performed according to 'MCC-1, Static Leach Test Procedure' in acid and basic solution. In this study, for the $Al_2O_3$-containing glasses, Na ion release from these glasses was higher in acid solution than in basic solution. As the content of $Fe_2O_3$ was increased in glasses, Na ion release was increased in acid solution, in spite of decrease of amount of total mass diminution.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.199-207
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• 2004

In order to develop a safe geological disposal concept for the HLW from the nuclear power plants in Korea, it is necessary to evaluate the safety of the disposal concept in an underground research tunnel in the same geological formation as the host rock mass. The design concept of a research tunnel depends on the actual disposal concept, repository geometry, experiments to be carried at the tunnel, and geological conditions. In this study, geological investigation had been carried out to develop the basic design of the small scale underground disposal research tunnel in KAERI.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.75-82
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• 2013

A large quantity of radioactive waste was generated during the decommissioning of the KRR and UCF. The radioactive waste was packed into 200 liter drums and 4m3 containers and these were temporarily stored onsite until their final disposal in the national repository facility. Some of the releasable waste was freely released and utilized for non-nuclear industries. The combustible wastes were treated by the utilization of an incinerator with a capacity of on average 20 kg/hr.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.283-289
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• 2003

In order to establish the licensing review system for LILW disposal facility, we have studied the licensing review structure of oversea's countries, including United State, Japan, and France. We have also reviewed the domestic licensing review structure and the current status of development of safety standards for LILW management. A licensing review for LILW disposal facility can be implemented in 5-6 steps according to Atomic Energy Act. It is estimated to take 32 months for the CP and the OL review for LILW disposal facility referencing to the licensing review practice of the nuclear power plant. To date, a total of 15 MOST Notices have been developed to apply to the safe management of radwaste and 5 more MOST Notices will be developed by 2005.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.303-313
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• 2023

Spent fuels (SFs) are stored in a storage pool after discharge from nuclear power plants. They can be transferred to for the further processes such as dry storage sites, processing plants, or disposal sites. One of important measures of SF is the burnup. Since the radioactivity of SF is strongly dependent on its burnup, the burnup of SF should be well estimated for the safe management, storage, and final disposal. Published papers about the methodology for the burnup estimation from the known activities of important radioactive sources are somewhat rare. In this study, we analyzed the dependency of the burnup on the important radiation source activities using ORIGEN-ARP, and suggested simple correlations that relate the burnup and the important source activities directly. A burnup estimation equation is suggested for PWR fuels relating burnup with total neutron source intensity (TNSI), initial enrichment, and cooling time. And three burnup estimation equations for major gamma sources, ¹³⁷Cs, ¹³⁴Cs, and ¹⁵⁴Eu are also suggested.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.876-883
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• 2023

Radioactive waste should be solidified before being disposed of in the repository to eliminate liquidity or dispersibility. Cement is a widely used solidifying media for radioactive waste, and cement solidified waste should satisfy the minimum compressive strength of the waste acceptance criteria of a radioactive repository. Although the compressive strength of waste should be measured by the test method provided by the waste acceptance criteria, the method differs depending on the operating repository of different countries. Considering the measured compressive strength changes depending on test conditions, the effect of test conditions should be analyzed to avoid overestimation or underestimation of the compressive strength during disposal. We selected test conditions such as the height-to-diameter ratio, loading rate, and porosity as the main factors affecting the compressive strength of cement solidified radioactive waste. Owing to the large variance in measured compressive strength, the effects of the test conditions were analyzed via statistical analyses using parametric and nonparametric methods. The results showed that the test condition of the lower loading rate, with a height-to-diameter ratio of two, reflected the actual cement content well, while the porosity showed no correlation. The compressive strength assessment method that reflects the large variance of strengths was suggested.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.227-234
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• 2019

A user-friendly modeling interface is developed for a process-based total system performance assessment framework (APro) specialized for a generic geological disposal system for high-level radioactive waste. The APro modeling interface is constructed using MATLAB, and the operator splitting scheme is used to combine COMSOL for simulation of multiphysics and PHREEQC for the calculation of geochemical reactions. As APro limits the modeling domain to the generic disposal system, the degree of freedom of the model is low. In contrast, the user-friendliness of the model is improved. Thermal, hydraulic, mechanical and chemical processes considered in the disposal system are modularized, and users can select one of multiple modules: "Default process" and multi "Alternative process". APro mainly consists of an input data part and calculation execution part. The input data are prepared in a single EXCEL file with a given format, and the calculation part is coded using MATLAB. The final results of the calculation are created as an independent COMSOL file for further analysis.

• Journal of Radiation Protection and Research
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• v.45 no.4
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• pp.178-186
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• 2020

Background: Reactor pressure vessel (RV) with internals (RVI) are activated structures by neutron irradiation and volume contaminated wastes. Thus, to develop safe and optimized disposal plan for them at a disposal site, it is important to perform exact activation calculation and evaluate the dose rate on the surface of casks which contain cut-off pieces. Materials and Methods: RV and RVI are subjected to neutron activation calculation via Monte Carlo methodology with MCNP6 and ORIGEN-S program-neutron flux, isotopic specific activity, and gamma spectrum calculation on each component of RV and RVI, and dose rate evaluation with MCNP6. Results and Discussion: Through neutron activation analysis, dose rate is evaluated for the casks containing cut-off pieces produced from decommissioned RV and RVI. For RV cut-off ones, the highest value of dose rate on the surface of cask is 6.97 × 10-1 mSv/hr and 2 m from it is 3.03 × 10-2 mSv/hr. For RVI cut-off ones, on the surface of it is 0.166 × 10-1 mSv/hr and 2 m from it is 1.04 × 10-1 mSv/hr. Dose rates for various RV and RVI cut-off pieces distributed lower than the limit except the one of 2 m from the cask surface of RVI. It needs to adjust contents in cask which carries highly radioactive components in order to decrease thickness of cask. Conclusion: Two types of casks are considered in this paper: box type for very-low-level waste (VLLW) as well as low-level waste (LLW) and cylinder type for intermediate-level waste (ILW). The results will contribute to the development of optimal loading plans for RV and RVI cut-off pieces during the decommissioning of nuclear power plant that can be used to prepare radioactive waste disposal plans for the different types of wastes-ILW, LLW, and VLLW.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.1
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• pp.63-72
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• 2009

Recently the importance and interest for the microbes has been increased because several important results for the effects of microbes on the radioactive waste disposal have been published continuously. In this study, research status and major results on the various roles and effects of microbes in the radioactive waste disposal have been investigated. We investigated and summarized the roles and major results of microbes in a multi-barrier system consisting of an engineered barrier and a natural barrier which is considered in radioactive waste disposal systems. For the engineered barrier, we discussed about the effects of microbes on the corrosion of a waste container and investigated the survival possibility and roles of microbes in a compacted bentonite buffer. For the natural barrier, the roles of microbes present in groundwaters and rocks were discussed and summarized with major results from natural analogue studies. Furthermore, we investigated and summarized the roles and various interactions processes of microbes and their effects on the radionuclide migration and retardation including recent research status. Therefore, it is expected that the effects and roles of microbes on the radioactive waste disposal can be rigorously evaluated if further researches are carried out for a long-term behavior of the disposal system in the deep geological environments and for the effects of microbes on the radionuclide migration through geological media.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.181-189
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• 2011

The smooth construction and operation of nuclear facilities requires understanding and support of both the local residents and the national people. It is essential that our country, which should maintain using nuclear energy for national energy security and economic growth, shall improve the social acceptance of nuclear energy. In order to identify the level of social acceptance of nuclear energy, this study investigated the perception of the local residents in Gyeongju and the public in other areas on a nuclear power plant and a low- and intermediate-level radioactive waste disposal facility through an individual interview. The subjects of the investigation were 450 persons. This study identified that perceptions of the respondents were somewhat dependent on the residential area, and derived the implications to be reflected in establishing the customized public-relation strategies.