• KSCE Journal of Civil and Environmental Engineering Research
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• v.41 no.2
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• pp.123-131
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• 2021

The buffer is a key component of an engineered barrier system that safeguards the disposal of high-level radioactive waste. Buffers are located between disposal canisters and host rock, and they can restrain the release of radionuclides and protect canisters from the inflow of ground water. Since considerable heat is released from a disposal canister to the surrounding buffer, the thermal conductivity of the buffer is a very important parameter in the entire disposal safety. For this reason, a lot of research has been conducted on thermal conductivity prediction models that consider various factors. In this study, the thermal conductivity of a buffer is estimated using the machine learning methods of: linear regression, decision tree, support vector machine (SVM), ensemble, Gaussian process regression (GPR), neural network, deep belief network, and genetic programming. In the results, the machine learning methods such as ensemble, genetic programming, SVM with cubic parameter, and GPR showed better performance compared with the regression model, with the ensemble with XGBoost and Gaussian process regression models showing best performance.

• The Journal of Engineering Geology
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• v.20 no.3
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• pp.271-279
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• 2010

Time-dependent behavior similar to secondary deformation related to mineral dissolution is easily observed when performing a laboratory pressure experiment. In this research, to observe the dissolution of quartz found in bentonite used as buffer material for the geological disposal of high-level waste (HLW) under conditions of high pH, we calculated the diffusion of $OH^-$ ions and the behavior of quartz dissolution using the homogenization analysis method. The results reveal that the rate of quartz dissolution is proportional to the temperature and interlayer water thickness. In particular, in a high-pH environment, the reacted area (and therefore the dissolution rate) increases with decreasing interlayer water thickness.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.4
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• pp.501-510
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• 2022

The nuclear criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of SNF disposal canister: the KBS-3 canister and small standardized transportation, aging and disposal (STAD) canister. Firstly, the KBS-3 & STAD canister containing four SNFs of the initial enrichment of 4.0wt% ²³⁵U and discharge burnup of 45,000 MWD/MTU were modelled. The k_eff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.79108, 0.78803, and 0.78484 & 0.76149, 0.75683, and 0.75444, respectively. Secondly, the KBS-3 & STAD canister with four SNFs of 4.5wt% and 55,000 MWD/MTU were modelled. The k_eff values for the cooling times of 40, 50, and 60 years were 0.78067, 0.77581, and 0.77335 & 0.75024, 0.74647, and 0.74420, respectively. Therefore, all cases met the performance criterion with respect to the k_eff value, 0.95. The STAD canister had the lower k_eff values than KBS-3. The neutron absorber plates in the STAD canister significantly affected the reduction in k_eff values although the distance among the SNFs in the STAD canister was considerably shorter than that in the KBS-3 canister.

• Tunnel and Underground Space
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• v.24 no.6
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• pp.405-417
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• 2014

In a high-level waste repository, the gap fill of the engineered barrier is an important component that influences the performance of the buffer and backfill. This paper reviewed the overseas status of R&D on the gap fill used engineered barriers, through which the concept of the gap fill, manufacturing techniques, pellet-molding characteristics, and emplacement techniques were summarized. The concept of a gap fill differs for each country depending on its disposal type and concept. Bentonite has been considered a major material of a gap fill, and clay as an inert filler. Gap fill was used in the form of pellets, granules, or a pellet-granule blend. Pellets are manufactured through one of the following techniques: static compaction, roller compression, or extrusion-cutting. Among these techniques, countries have focused on developing advanced technologies of roller compression and extrusion-cutting techniques for industrial pellet production. The dry density and integrity of the pellet are sensitive to water content, constituent material, manufacturing technique, and pellet size, and are less sensitive to the pressure applied during the manufacturing. For the emplacement of the gap fill, pouring, pouring and tamping, and pouring with vibration techniques were used in the buffer gap of the vertical deposition hole; blowing through the use of shotcrete technology and auger placement and compaction techniques have been used in the gap of horizontal deposition hole and tunnel. However, these emplacement techniques are still technically at the beginning stage, and thus additional research and development are expected to be needed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.217-224
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• 2008

In deep geological disposal system, the integrity of a disposal canister having spent fuels is very important factor to assure the safety of the repository system. This disposal canister is one element of the engineered barriers to isolate and to delay the radioactivity release from human beings and the environment for a long time so that the toxicity does not affect the environment. The main requirement in designing the deep geological disposal system is to keep the buffer temperature below 100$^{\circ}C$ by the decay heat from the spent fuels in the canister in order to maintain the integrity of the buffer material. Also, the disposal canister can endure the hydraulic pressure in the depth of 500 m and the swelling pressure of the bentonite as a buffer. In this study, new concept of the disposal canister for the CANDU spent fuels which were considered to be disposed without any treatment was developed and the thermal stability and the structural integrity of the canister were analysed. The result of the thermal analysis showed that the temperature of the buffer was 88.9$^{\circ}C$ when 37 years have passed after emplacement of the canister and the spacings of the disposal tunnel and the deposition holes were 40 m and 3 m, respectively. In the case of structural analysis, the result showed that the safety factors of the normal and the extreme environment were 2.9 and 1.33, respectively. So, these results reveal that the canister meets the thermal and the structural requirements in the deep geological disposal system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.105-112
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• 2004

A waste container, one of the key components of a multi-barrier system in a potential high level radioactive waste (HLW) repository in Korea ensures the mechanical stability against the lithostatic pressure of a deep geologic medium and the swelling pressure of the bentonite buffer. Also, it delays potential release of radionuclides for a certain period of time, before it is corroded by intruding impurities. Even though the material of a waste container is carefully chosen and its manufacturing processes are under quality assurance processes, there is a possibility of initial defects in a waste container during manufacturing. Also, during the deposition of a waste container in a repository, there is a chance of an incident affecting the integrity of a waste container. In this study, the appropriate Features, Events, and Processes(FEP's) to describe these incidents and the associated scenario on radionuclide release from a container to the biosphere are developed. Then the total system performance assessment on the Initial waste Container Failure (ICF) scenario was carried out by the MASCOT-K, one of the probabilistic safety assessment tools KAERI has developed. Results show that for the data set used in this paper, the annual individual dose for the ICF scenario meets the Korean regulation on the post closure radiological safety of a repository.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.229-232
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• 2003

A waste container, one of the key compartments in a multi-barrier system for a potential high level radioactive waste (HLW) repository in Korea ensures the mechanical stability against the lithostatic pressure of a deep geologic strata and the swelling pressure of the bentonite buffer. Also, it prohibits potential release of radionuclides for a certain period of time. before it is corroded by impurities. Even though the materials of a waste container is carefully chosen and all manufacturing processes are under heavy quality assurance, there might be a slight chance of intial defects in a waste container. Also, during the deposition of a waste container in a repository, there might be a chance of an incident affecting the integrity of a waste container. In this study, the FEP's and the scenarios over radiological impact of a potential initial waste container defect was developed. Then the total system performance assessment on this initial waste container failure (ICF) scenario was carried out by the MASCOT-K, one of the probabilistic safety assessment tools KAERI has developed. Results show that for the data set studied in this paper, the annual individual dose by the ICF scenario well meets the KINS regulation.

• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.29-40
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• 2013

Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel) for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.1
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• pp.37-43
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• 2014

A parametric sensitivity to the annual exposure dose rate to the farming exposure group has been deterministically carried out for three principal elements identified in the near-field of the pyroprocessed waste repository system as a series study of Part 1 of the coupled paper with the same title. Credit time for both metal and ceramic containers, annual nuclide release rete and the degree of loss of bentonite buffer around the container are selected and investigated deterministically for important nuclides. To this end the A-KRS has been assessed and then compared among each other with the normal, the worst, and the best case scenarios associated with their extreme values these elements could have. All the elements are shown to be sensitive to the results as was in Part 1. Methodology studied through this study and the results are expected to make a good feedback to the repository design.

• Tunnel and Underground Space
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• v.34 no.1
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• pp.71-87
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• 2024

In deep geological disposal of high-level radioactive waste, the surrounding rock at the immediate vicinity of the deposition hole may experience localized changes in permeability due to in-situ stress at depth, swelling pressure from resaturated bentonite buffer, and the heat generated from the decay of radioactive isotopes. In this study, experimental data on changes in permeability of granite, a promising candidate rock type in South Korea, were obtained by applying various confining pressures and temperature conditions expected in the actual disposal environment. By conducting the permeability test on KURT granite specimens under three or more hydrostatic pressure conditions, the relation in which the permeability decreases exponentially as the confining pressure increases was derived. The temperature-induced changes in permeability were found to be negligible at temperatures below the expected maximum of 90℃. In addition, by establishing a relation in which the initial permeability is proportional to the power of the initial porosity, it was possible to estimate permeability value for granite with a specific porosity under a certain confining pressure.