• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• 제6권1호
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• pp.45-54
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• 2008

A new proposed repository has a final capacity of 800,000 drums radioactive waste. Most of foreign repositories have a general practice of segregating control zones which mainly contributes to classification of degree of control, whether it is called buffer zone or not. Domestic regulatory requirements of establishment of buffer zone in a repository are not much different from those of nuclear power plants for operation period, in which satisfactory design objective or performance objective is the most important factor in determination of the buffer zone. The meaning of buffer zone after closure is a minimum requested area which can prevent inadvertant intruders from leading to non-allowable exposure during institutional control period. Safety assessment with drinking well scenario giving rise to the highest probability of exposure among the intruder's actions can verify fulfillment of the buffer zone which is determined by operational safety of the repository. At present. for the repository to be constructed in a few years, the same procedure and concept as described in this paper are applied that can satisfy regulatory requirements and radiological safety as well. However, the capacity of the repository will be stepwise extended upto 800,000 drums, consequently its layout will be varied too. Timely considerations will be necessary for current boundary of the buffer zone which has been established on the basis of 100,000 drums disposal.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• 제15권1호
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• pp.45-63
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• 2017

The oldest commercial reactor in South Korea, Kori-1 Nuclear Power Plant (NPP), will be shut down in 2017. Proper treatment for decommissioning wastes is one of the key factors to decommission a plant successfully. Particularly important is the recycling of clearance level or very low level radioactively contaminated metallic wastes, which contributes to waste minimization and the reduction of disposal volume. The aim of this study is to introduce a conceptual design of a recycle system and to evaluate the doses incurred through defined work flows. The various architecture diagrams were organized to define operational procedures and tasks. Potential exposure scenarios were selected in accordance with the recycle system, and the doses were evaluated with the RESRAD-RECYCLE computer code. By using this tool, the important scenarios and radionuclides as well as impacts of radionuclide characteristics and partitioning factors are analyzed. Moreover, dose analysis can be used to provide information on the necessary decontamination, radiation protection process, and allowable concentration limits for exposure scenarios.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• 제3권4호
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• pp.341-348
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• 2005

Bentonite has been considered as a candidate buffer material in the underground repository for the disposal of high-level radioactive waste because of its low permeability, high sorption capacity, self sealing characteristics, and durability in nature. In this study, the potential for separation of bentonite particles caused by the groundwater erosion was studied experimentally for a Korean Ca-bentonite under the relevant repository conditions. Results showed that bentonite particles can be generated at the bentonite/granite interface and mobilized by the water flow although the intrusion of bentonite into fracture by swelling pressure was observed to be small. Different processes of mobilization of theses colloids from the compacted bentonite block have been identified in this study. The concentration of particles eluted in water was increased as the flow rate increased. Thus the result reveals that the erosion of the bentonite surface due to the groundwater flow together with intrusion processes is the main mechanism that can mobilize bentonite colloids in the fracture of the granite.

• Nuclear Engineering and Technology
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• 제25권1호
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• pp.63-70
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• 1993

A simplified safety assessment is carried out on rock-cavern type disposal of LLW using the analytical repository source term (REPS) model. For reliable prediction of the leach rates for various radionuclides, degradation of concrete structures, corrosion rate of waste container, degree of corrosion on the container surface, and the characteristics of radionuclides are considered in the REPS model. The results of preliminary assessment show that Cs-137, Ni-63, and Sr-90 are dominant. For the parametric uncertainty and sensitivity analysis, Latin hypercube sampling technique and rank correlation technique are applied. The results of the potential public health impacts show that radiological dose to intruder in the worst case scenario will be negligible and that more attention should be given to near-field performance.

• Journal of Radiation Protection and Research
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• 제16권1호
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• pp.43-52
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• 1991

A comparison study is performed on dose assessment programs including DOSE, LIMCAL, and PABLM, DOSE is a program developed for preliminary safety assessments of the low- and intermediate- level radioactive waste disposal, and the others are existing programs applicable to similar calculations. The results show acceptable agreement within an order of magnitude(mrem/yr) except for C-14 and Pu-239. PABLM results higher dose for C-14 while lower value for Pu-239 in comparison with those from DOSE or LIMCAL. It is found that the discrepancy in C-14 is due to difference in transport model introduced and that in Pu-239 is from the different value of dose conversion factor to each program.

• Journal of the Korean Society of Systems Engineering
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• 제17권1호
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• pp.11-20
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• 2021

Decontamination of systems, structures and components (SSC) during the decommissioning of a Nuclear Power Plant (NPP) can be for a variety of reasons. The main reasons for decontamination are: to reduce the contamination of SSC to a reasonably low level, to reduce the potential for the spread of contaminants into the environment and to reduce the cost of disposal due to the reduced level of contamination in a particular SSC. The decontamination technique can be aggressive or non-aggressive depending on the intent after the decontamination process. Aggressive decontamination technique is used when the intent is not to reuse the SSC while a non-aggressive decontamination technique is used with the intent of SSC reuse. For different SSCs there are different decontamination techniques that can be used, each having its own advantages and drawbacks. Metal components such as pipes in the nuclear power plant account for a large amount of nuclear wastes generated. Some of these wastes can be reused if the contaminant level is reduced to an acceptable level. Laser ablation is a non-aggressive decontamination technique that can be used to reduce the contamination in pipes to an acceptable level with no secondary waste generated during the process. The operation and control of a laser ablation device must be precise to achieve a high decontamination factor. This precision can be achieved by a well-designed motion control system. For this purpose, a motion control system was developed consisting of two parts: the first part being the precise control of the laser ablation device inside the pipe and the second part is the control of the laser ablation device outside the pipe. This paper describes the Systems Engineering approach for the development process of a motion control system for the Laser decontamination system.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 한국방사성폐기물학회 2004년도 학술논문집
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• pp.419-424
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• 2004

Various radiation counters have been using to determine radioactivity of radwastes for disposal. A radiation counting system was set up using a radiation detector chosen in this study and its stability was investigated through the periodic determination of background and counting efficiencies in accordance with a quality control program to increase the confidence level. The average background level for the $\gamma$-spectrometer was 1.59 cps and the average counting level for the standard sample was 45248 Ops within $2{\sigma}$ confidence levels. The average alpha background level for the low background ${\alpha}{\beta}$ counting system was 0.31 cpm and the efficiency for alpha counting was 34.38%. The average beta background level for the ${\alpha}{\beta}$ counting system was 1,30 cpm and the efficiency for beta counting was 46.5%, The background level in the region of 3H and 14C for the liquid scintillation counting system was 2.52 and 3.31 cpm and the efficiency for alpha counting was 58.5 and 95.6%, respectively. The minimum detectable activity for the$\gamma$-spectrometer was found to be 3.2 Bq/$m\ell$ and 3.8 Bq/$m\ell$ for the liquid scintillation counter, and 20.5 and 23.0 Bq/$m\ell$, respectively for the $\alpha$ and $\beta$ counting system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• 제10권4호
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• pp.247-254
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• 2012

Vitrification technology has been widely applied as one of effective processing methods for wastes generated in nuclear power plants. The advantage of vitrifying for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. Recently, a filter using on HVAC(Heating Ventilating & Air Conditioning System) is composed with media (glass fiber) and separator (aluminum film) has been studied the proper treatment technology for meeting the waste disposal requirement. Present paper is a feasibility study for the filter vitrification that developing of the glass compositions for filter melting and melting test for physicochemical characteristic evaluation. The aluminum metal of film type is preparing with 0.5 cm size for proper mixing with glass frit, glass fiber is also preparing with 1 cm size within crucible. The glass compositions should be developed considering molten glass are related with wastes reduction. Glass compositions obtained from developing on glass formulation are mainly composed of $SiO_2$ and $B_2O_3$ for aluminum metal. A variety of factors obtained from the glass formulation and melting test are reviewed, which is feeding rate and glass characteristics of final products such as durability for implementing the wastes disposal requirement.

• Journal of the Computational Structural Engineering Institute of Korea
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• 제34권5호
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• pp.339-345
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• 2021

Finite element analysis of the silo type underground opening for low- and intermediate-level radioactive waste (LILW) disposal facilities in Korea is presented in this study. The silo wall is circular and the roof is made up of domes. The silo wall is 25 meters in diameter, 35 meters in height, and the dome is 30 meters in diameter and 17.4 meters in height, and it is located at -80 meters to -130 meters at sea level. Although six silos have been constructed in the first stage and are in operation, only one silo was considered in this study. The two-dimensional axial symmetric finite element model, as well as the three-dimensional finite element model were made using the computer program SMAP-3D. Generalized Hoek and Brown Model was used for the numerical analyses. The finite element analysis of the silo type underground opening was carried out under various lateral pressure coefficients (defined as ratio of average horizontal to vertical in-situ stress), and the numerical results of these analyses were examined.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• 제6권4호
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• pp.265-282
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• 2008

Based on the site characterization works in a low and intermediate level waste(LILW) repository site, the numerical simulations for groundwater flow were carried out in order to understand the groundwater flow system of repository site. To accomplish the groundwater flow modeling in the repository site, the discrete fracture network(DFN) model was constructed using the characteristics of fracture zones and background fractures. At result, the total 10 different hydraulic conductivity(K) fields were obtained from DFN model stochastically and K distributions of constructed mesh were inputted into the 10 cases of groundwater flow simulations in FEFLOW. From the total 10 numerical simulation results, the simulated groundwater levels were strongly governed by topography and the groundwater fluxes were governed by locally existed high permeable fracture zones in repository depth. Especially, the groundwater table was predicted to have several tens meters below the groundwater table compared with the undisturbed condition around disposal silo after construction of underground facilities. After closure of disposal facilities, the groundwater level would be almost recovered within 1 year and have a tendency to keep a steady state of groundwater level in 2 year.