• Journal of the Korean Geotechnical Society
• /
• v.37 no.4
• /
• pp.19-26
• /
• 2021

The high-level radioactive waste (HLW) produced from nuclear power plants is disposed in a rock-mass at a depth of hundreds meters below the ground level. Since HLW is very dangerous to human being, it must be disposed of safely by the engineered barrier system (EBS). The EBS consists of a disposal canister, backfill material, buffer material, and so on. When the components of EBS are installed, gaps inevitably exist not only between the rock-mass and buffer material but also between the canister and buffer material. The gap can reduce water-retarding capacity and heat release efficiency of the buffer material, so it is necessary to investigate properties of gap-filling materials and to analyze gap spacing effect. Furthermore, there has been few researches considering domestic disposal system compared to overseas researches. In this reason, this research derived the peak temperature of the bentonite buffer material considering domestic disposal system based on the numerical analysis. The gap between the canister and buffer material had a minor effect on the peak temperature of the bentonite buffer material, but there was 40% difference of the peak temperature of the bentonite buffer material because of the gap existence between the buffer material and rock mass.

• Tunnel and Underground Space
• /
• v.34 no.1
• /
• pp.28-53
• /
• 2024

In general, high-level radioactive waste (HLW) generated as a result of nuclear power generation should be disposed within the country. Determination of the disposal site and host rock for HLW deep geological repository is an important issue not only scientifically but also politically, economically, and socially. Considered host rock types worldwide for geological disposal include crystalline rocks, sedimentary rocks, volcanic rocks, and salt dome. However, South Korea consists of various rock types except salt dome. This paper not only analyzed the geological and rock mechanical characteristics on a nationwide scale with the preliminary results on various rock type studies for the disposal host rock, but also reviewed the characteristics and possibility of various rock types as a host rock through deep drilling surveys. Based on the nationwide screening for host rock types resulted from literature review, rock distributions, and detailed case studies, Jurassic granites and Cretaceous sedimentary rocks (Jinju and Jindong formations) were derived as a possible candidate host rock types for the geological disposal. However, since the analyzed data for candidate rock types from this study is not enough, it is suggested that the disposal rock type should be carefully determined from additional and detailed analysis on disposal depth, regional characteristics, multidisciplinary investigations, etc.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2019.05a
• /
• pp.169-170
• /
• 2019

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.2
• /
• pp.197-207
• /
• 2006

The engineering performance of a high level waste repository is significantly dependent upon the T-H-M behavior in the engineered barrier system. An engineering-scale test facility (KENTEX) was set up to validate the T-H-M behaviors in the buffer of a reference disposal system developed in the 2002. The validation tests started on May 31, 2005 and is now in progress. The KENTEX facility and validation test programme are introduced, and pre-operation calculations are also presented to give information on the sensitive location of sensors and operational conditions. This test will provide information (e.g., large-scale apparatus, sensors, monitoring system etc.) needed for 'in-situ' tests, make the validation of a T-H-M model for the T-H-M performance assessment of the reference disposal system, and demonstrate the engineering feasibility of fabricating and emplacing the buffer of a repository.

• Tunnel and Underground Space
• /
• v.11 no.2
• /
• pp.132-145
• /
• 2001

The purpose and need of the study is to quantify the advantage or disadvantage of the environmental friendliness of the partitioning of nuclear fuel cycle. To this end, a preliminary study on the quantitative effect of the partition on the permanent disposal of spent PWR and CANDU fuel (HLW) was carried out. Before any analysis, the so-called reference radionuclide release scenario from a potential repository embedded into a crystalline rock was developed. Firstly, the feature, event and processes (FEPs) which lead to the release of nuclides from waste disposed of in a repository and the transport to and through the biosphere were identified. Based on the selected FEPs, the ‘Well Scenario’which might be the worst case scenario was set up. For the given scenario, annual individual doses to a local resident exposed to radioactive hazard were estimated and compared to that from direct disposal. Even though partitioning and transmutation could be an ideal solution to reduce the inventory which eventually decreases the release time as well as the peaks in the annual dose and also minimize the repository area through the proper handling of nuclides, it should overcome major disadvantages such as echnical issues on the partitioning and transmutation system, cost, and public acceptance, and environment friendly issues. In this regard, some relevant issues are also discussed to show the direction for further studies.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.3 no.4
• /
• pp.349-358
• /
• 2005

In this study, analysis of the disposal tunnel spacing and disposal pit pitch was carried out, as a factor of the design to estimate the scale and layout of the repository. To do this, based on the reference repository concept and the engineered barrier concept, several cross sections of the disposal tunnel and disposal pit were established. After then, the mechanical and thermal stabilities of the established tunnels were analyzed. Also, an optimized disposal tunnel spacing and the disposal pit pitch reducing the excavation volume was proposed. The results of these analyses can be used in the deep geological repository design. The detailed analyses by the exact site characteristics data to reduce the uncertainty of the site and the modification for the optimization are required.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2003.11a
• /
• pp.229-232
• /
• 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
• /
• v.45 no.1
• /
• pp.29-40
• /
• 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)
• /
• v.2 no.4
• /
• pp.253-262
• /
• 2004

Construction of tunnels in a deep crystalline host rock for a potential High-Level Radioactive Waste(HLW) repository inevitably generates an excavation disturbed zone (EDZ). There have been a series of debates on whether a permeability in an EDZ increases or not and what would be the maximum depth of an EDZ. Recent studies show mixed opinions on permeability. However, there has been an international consensus on the thickness of an EDZ; 30 cm for TBM and 1 meter for controlled blast. One of the impacts of an EDZ is on determining the distance between adjacent deposition holes. The void gap by the excavation hinders relaxation of temperature profiles so that the current Korean reference designing distance between holes should be stretched out more to keep the maximum temperature in a buffer region below 100 degrees Celsius. The other impact of an EDZ is on the long-term post closure radiological safety. To estimate the impact, the reference scenario, the well scenario, is chosen. Released nuclides diffuse through a bentonite buffer region experiencing strong sorption and reach a fracture surrounded by a porous medium. Inside a fractured porous region, radionuclides migrate by advection and dispersion with matrix diffusion into a porous medium. Finally, they reach a well assumed to be a source of potable water for local residents. The annual individual dose is assessed on this well scenario to find out the significance of an EDZ. A profound sensitivity study was performed, but all results show that the impact is negligible. Even though the role of an EDZ turns out to be limited on overall safety assessment, still it is worthwhile to study the chemical role of an EDZ, such as a potential source for natural colloids, potential sealing of an open fracture by fine clay particles generated by the process of an EDZ, and alteration of a sorption mechanism by an EDZ in the future.

• Nuclear Engineering and Technology
• /
• v.36 no.2
• /
• pp.196-202
• /
• 2004

KAERI (Korea Atomic Energy Research Institute) is at present preparing a preliminary performance assessment to set up the HLW disposal concept of Korea. The solubility of the radionuclides contained in HLW is necessary as a source term in order to predict their potential migration in both the near and far fields. The solubility of actinides (Th, Am, U, Np and Pu) for a reference deep groundwater of Korea has been calculated using a geochemical code with thermodynamic data selected by a peer review of existing thermodynamic databases and literature. The solubilities from the experimental study and/or field observations from natural analogue studies are compared. The sensitivity of solubility to the variability of three main parameters of groundwater (pH, Eh, and carbonate concentration) is also investigated. The results of the sensitivity analysis show that the solubility of actinides strongly depends on the parameters considered. Within the range of parameter values studied (pH=7 to 10, Eh=-0.4 to -0.1V, and carbonate concentration=1.E-5 to 1.E-2 mol/L), the solubility of each actinide exists between 1.4E-10 and 1.6E-6 mol/L for Am, 4.9E-9 and 2.8E-6 mol/L for Th, 3.2E-9 and 5.7E-4 mol/L for U, 1.1E-9 and 1.0E-7 mol/L for Np, and 4.0E-11 and 2.8E-6 mol/L for Pu, respectively.