• Journal of Radiation Industry
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• v.18 no.1
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• pp.43-52
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• 2024

Currently, there are no interim storage facilities and permanent disposal facilities in Korea, so all spent nuclear fuels are temporarily stored. However, the temporary storage facility is approaching saturation, and as a measure to this, the 2nd Basic Plan for the Management of High-Level Radioactive Waste presented an operation plan for dry interim storage facilities and dry temporary storage facilities on the NPP on-site. The dry storage can be operated in various ways, and to select the optimal dry storage method, the reduction of exposure for workers must be considered. Accordingly, it is necessary to develop a worker exposure scenario according to the dry storage method and evaluate and compare the radiological impact for each method. The purpose of this study is to develop an exposure scenario for workers transporting spent nuclear fuel by dry storage method. To this end, first, the operation procedure of the foreign commercial spent nuclear fuel dry storage system was analyzed based on the Final Safety Analysis Report (FSAR). 1) the concrete overpack-based system, 2) the metal overpack-based system, and 3) the vertical storage module-based system were selected for analysis. Factors were assumed that could affect the type of work (working distance, working hours, number of workers, etc.) during transportation work. Finally, the work type of the processes involved in transporting spent nuclear fuel by dry storage method was set, and an exposure scenario was developed accordingly. The concrete overpack method, the metal overpack method, and the vertical storage module method were classified into a total of 31, 9, and 23 processes, respectively. The work distance, work time, and number of workers for each process were set. The product of working hours and number of workers (Man-hour) was set high in the order of concrete overpack method, vertical storage module method, and metal overpack method, and short-range work (10 cm) was most often applied to the concrete overpack method. The results of this study are expected to be used as basic data for performing radiological comparisons of transport workers by dry storage method of spent nuclear fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.219-228
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• 2012

A long-term R&D program for HLW disposal technology development was launched in 1997 in Korea and Korea Reference disposal System(KRS) for spent fuels had been developed. After then, a recycling process for PWR spent fuels to get the reusable material such as uranium or TRU and to reduce the volume of radioactive waste, called Pyro-process, is being developed. This Pyro-process produces several kinds of wastes including metal waste and ceramic waste. In this study, the characteristics of the waste from Pyro-process and the concepts of a disposal container for the wastes were described. Based on these concepts, thermal analyses were carried out to determine a layout of the disposal area of the ceramic wastes which was classified as a high level waste and to develop the disposal system called A-KRS. The location of the final repository for A-KRS is not determined yet, thus to review the potential repository domains, the possible layout in the geological characteristics of KURT facility site was proposed. These results will be used in developing a repository system design and in performing the safety assessment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.313-319
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• 2019

The A-KRS (Advanced Korean Reference Disposal System) is the disposal concept for pyroprocessed waste, which has been developed by the Korea Atomic Energy Research Institute. In this disposal concept, the amount of high-level radioactive waste is minimized using pyrochemical process, called pyroprocessing. The produced pyroprocessed waste is then solidified in the form of monazite ceramic. The final product of ceramic wastes will be disposed of in a deep geological repository. By the way, the decay heat is generated due to the radioactive decay of fission products and raises the temperature of buffer materials in the near field of radioactive waste repository. However, the buffer temperature must be kept below $100^{\circ}C$ according to the safety regulation. Usually, the temperature can be controlled by variation of the canister interdistance. However, KAERI has modelled thermal analysis under the boundary condition, where the waste canisters are in direct contact with each other. Therefore, a reliable temperature analysis in the disposal system may fail because of unknown thermal resistence values caused by the spatial gap between waste canisters. In the present work, we have performed thermal analyses considering the gap between heating elements and canisters at the beginning of canister loading into the radioactive waste repository. All thermal analyses were performed using the COMSOL software package.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.110-116
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• 2020

As the management plan for domestic spent nuclear fuel is delayed, the storage of the operating nuclear power plant is approaching saturation, and the Kori 1 Unit that has reached its end of operation life is preparing for the dismantling plan. The first stage of dismantling is the transfer of spent nuclear fuel stored in storage at plants. The spent fuel management process leads to temporary storage, interim storage, reprocessing and permanent disposal. In this paper, the technical issues to be considered when transporting spent fuel in this process are summarized. The spent fuels are treated as high-level radioactive waste and strictly managed according to international regulations. A series of integrity tests are performed to demonstrate that spent fuel can be safely stored for decades in a dry environment before being transferred to an intermediate storage facility. The safety of spent fuel transport container must be demonstrated under normal transport conditions and virtual accident conditions. IAEA international standards are commonly applied to the design of transport containers, licensing regulations and transport regulations worldwide. In addition, each country operates a physical protection system to reduce and respond to the threat of radioactive terrorism.

• Nuclear Engineering and Technology
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• v.34 no.3
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• pp.218-231
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• 2002

Using a three-dimensional numerical code, B3R developed for nuclide transport of an arbitrary length of decay chain in the buffer between the canister and adjacent rock in a high- level radioactive waste repository by adopting a finite difference method utilizing the control- volume scheme, some illustrative calculations have been done. A linear sorption isotherm, nuclide transport due to diffusion in the buffer and the rock matrix, and advection and dispersion along thin rigid parallel fractures existing in a saturated porous rock matrix as well as diffusion through the fracture wall into the matrix is assumed. In such kind of repository, buffer and rock matrix are known to be important physico-chemical harriers in nuclide retardation. To show effects of buffer and rock matrix on nuclide transport in HLW repository and also to demonstrate usefulness of B3R, several cases of breakthrough curves as well as three- dimensional plots of concentration isopleths associated with these two barriers are introduced for a typical case of decay chain of $^{234}$ Ulongrightarrow$^{230}$ Thlongrightarrow$^{226}$ Ra, which is the most important chain as far as the human environment is concerned.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.274-279
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• 1996

Diffusion coefficient is a critical parameter for predicting radiological source term(migration rate and flux of radionuclide) through given near field conditions in spent fuel or high level waste repository. The effect of exchangeable cation-$Na^+$ and $Ca^{2+} - on the diffusion of $I^- \;and^3H$ (as HTO) in compacted bentonite was examined using a through-diffusion method. Bentonite material used here was compacted to a density of 1.3 Mg/m$^3$, and Na-bentonite was saturated with a solution of 100 mol NaCl/m$^3$ and Ca-bentonite with 50 $mol\;CaCl_2$/m$^3$. The results show that effective diffusion coefficients are generally higher by a factor of two to five in Ca-than Na-clay. This is attributed to the larger particle size of Ca-compared to Na-bentonite; hence, Ca-bentonite has a greater proportion of relatively large pores, which make a greater contribution to mass transport than small pores. Although the nature of the exchangeable cation affects mass diffusion in compacted bentonite, the effect is small and not likely to influence performance assessment modeling of compacted bentonite-based barriers.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4474-4480
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• 2022

Deep geological disposal using a multibarrier system is a promising solution for treating high-level radioactive (HLRW) waste. The HLRW canister represents the first barrier for the migration of radionuclides into the biosphere, therefore, the corrosion behavior of canister materials is of significance. In this study, the electrochemical behaviors of SS316L, Ti-Gr.2, Alloy 22, and Cu in naturally aerated KAERI underground research tunnel (KURT) groundwater solutions were examined. The corrosion potential, current, and impedance spectra of the test materials were recorded using electrochemical methods. According to polarization and impedance measurements, Cu exhibits relatively higher corrosion rates and a lower corrosion resistance ability than those exhibited by the other materials in the given groundwater condition. In the anodic dissolution tests, SS316L exposed to the groundwater solution exhibited the most uniform corrosion, as indicated by its surface roughness. This phenomenon could be attributed to the extremely low concentration of chloride ions in KURT groundwater.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.1
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• pp.13-21
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• 2011

To characterize the site specific properties of a study area for high-level radioactive waste disposal research in KAERI, the several geological investigations such as surface geological surveys and borehole drillings were carried out since 1997. Especially, KURT (KAERI Underground Research Tunnel) was constructed to understand the further study of geological environments in 2006. As a result, the first geological model of a study area was constructed by using the results of geological investigation. The objective of this research is to construct a hydrogeological model around KURT area on the basis of the geological model. Hydrogeological data which were obtained from in-situ hydraulic tests in the 9 boreholes were estimated to accomplish the objective. And, the hydrogeological properties of the 4 geological elements in the geological model, which were the subsurface weathering zone, the log angle fracture zone, the fracture zones and the bedrock were suggested. The hydrogeological model suggested in this study will be used as input parameters to carry out the groundwater flow modeling as a next step of the site characterization around KURT area.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.279-291
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• 2010

To characterize geological features in study area for high-level radioactive waste disposal research, KAERI (Korea Atomic Energy Research Institute) has been performing the several geological investigations such as geophysical surveys and borehole drilling since 1997. Especially, the KURT (KAERI Underground Research Tunnel) constructed to understand the deep geological environments in 2006. Recently, the deep borehole of 500 m depths was drilled to confirm and validate the geological model at the left research module of the KURT. The objective of this research was to identify the geological structures around KURT using the data obtained from the deep borehole investigation. To achieve the purpose, several geological investigations such as geophysical and borehole fracture surveys were carried out simultaneously. As a result, 7 fracture zones were identified in deep borehole located in the KURT. As one of important parts of site characterization on KURT area, the results will be used to revise the geological model of the study area.

• Journal of the Mineralogical Society of Korea
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• v.26 no.2
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• pp.119-127
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• 2013

Borosilicate glass dissolution is an important chemical process that impacts the glass durability as nuclear waste form that may be used for high-level radioactive waste disposal. Experiments reported that the glass dissolution rates are strongly dependent on the bulk composition. Because some relationship exists between glass composition and molecular-structure distribution (e.g., non-bridging oxygen content of $SiO_4$ unit and averaged coordination number of B), the composition-dependent dissolution rates are attributed to the bulk structural changes corresponding to the compositional variation. We examined Na borosilicate glass structures by performing classical molecular dynamics (MD) simulations for four different chemical compositions ($xNa_2O{\cdot}B_2O_3{\cdot}ySiO_2$). Our MD simulations demonstrate that glass surfaces have significantly different chemical compositions and structures from the bulk glasses. Because glass surfaces forming an interface with solution are most likely the first dissolution-reaction occurring areas, the current MD result simply that composition-dependent glass dissolution behaviors should be understood by surface structural change upon the chemical composition change.