• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2576-2581
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• 2021

The thermal stability and crystallization behaviors of La₂O₃ containing B₂O₃-CaO-Al₂O₃ glass waste forms were investigated to evaluate the stability of waste form during emergencies in deep geological disposal. For glasses containing 15% La₂O₃, LaBO₃ phases were observed as major crystals from 780 ℃ and exhibited needlelike structures. Al, Ca, and O were homogeneously distributed throughout the entire specimen, while some portions of B and La were concentrated in some parts. By differential thermal analysis at various heating rates, the activation energy for grain growth and the crystallization rate of LaBO₃ were calculated to be 12.6 kJ/mol and 199.5 kJ/mol, respectively. These values are comparable to other waste forms being developed for the same purpose.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1616-1622
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• 2022

Selenium has been identified as an element of interest for the safety assessment of a deep geological repository (DGR) for used nuclear fuel. In Canada, groundwaters at DGR depths in sedimentary rocks have been observed to have a high ionic strength. This paper examines the sorption behavior of Se(-II) onto illite, MX-80 bentonite, Queenston shale, and argillaceous limestone in Na-Ca-Cl solutions of varying ionic strength (0.1-6 mol/kgw (m)) and across a pH range of 4-9. Little ionic strength dependence for Se(-II) sorption onto all solids was observed except that sorption at high ionic strength (6 m) was generally slightly lower than sorption at low ionic strength (0.1 m). Illite and MX-80 exhibited the expected results for anion sorption, while shale and limestone exhibited more constant sorption across the pH range tested. A non-electrostatic surface complexation model successfully predicted sorption of Se(-II) onto illite and MX-80 using the formation of an inner-sphere surface complex and an outer-sphere surface complex. Optimized values for the formation reactions of these surface species were proposed.

• Journal of Korean Tunnelling and Underground Space Association
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• v.26 no.3
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• pp.191-208
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• 2024

In this study, a hydro-mechanical-damage coupled analysis model was developed to evaluate the structural safety of radioactive waste disposal structures. The Mazars damage model, widely used to model the fracture behavior of brittle materials such as rocks or concrete, was coupled with conventional hydro-mechanical analysis and the developed model was verified via theoretical solutions from literature. To derive the numerical input values for damage-coupled analysis, uniaxial compressive strength and Brazilian tensile strength tests were performed on concrete samples made using the mix ratio of the disposal concrete silo cured under dry and saturated conditions. The input factors derived from the laboratory-scale experiments were applied to a two-dimensional finite element model of the concrete silos at the Wolseong Nuclear Environmental Management Center in Gyeongju and numerical analysis was conducted to analyze the effects of damage consideration, analysis technique, and waste loading conditions. The hydro-mechanical-damage coupled model developed in this study will be applied to the long-term behavior and stability analysis of deep geological repositories for high-level radioactive waste disposal.

• Journal of the Mineralogical Society of Korea
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• v.27 no.4
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• pp.293-299
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• 2014

The bentonite, a buffer material, is essential for the deep geological disposal of HLW (high-level radioactive waste), and it is important to know its characteristic long-term evolution in the underground environment. With an assumption that the concentration of aqueous copper ions will increase if copper-coated materials on a metal canister are corroded, we examined some characteristic ion-exchanges and cation release phenomena occurring in the bentonite clay (montmorillonite) interacted with aqueous Cu cations. During the interaction between dissolved copper and bentonite, Na rather than Ca cations in the expandable clay were preferentially replaced by Cu ions in the experiment. In addition, the Cu-exchanged montmorillonite was characterized by an asymmetric X-ray diffracted pattern with relatively collapsed interlayers compared to the raw sample. These results indicate that the gradual change of the original bentonite property may occur in a underground disposal condition. We are going to further study the characteristic chemical and mineralogical changes of the bentonite buffer to be used for the disposal site by conducting additional experiments.

• Tunnel and Underground Space
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• v.33 no.1
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• pp.10-28
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• 2023

Long-term shear behavior of the rock discontinuities should be analyzed and its stability should be evaluated to ensure the long-term stability of a high-level radioactive waste disposal repository. The long-term shear behavior of the discontinuities can be modeled with creep and RSF models. The shear creep test, velocity step test, and slide-hold-slide test can be performed to determine their model parameters or analyze the shear behavior by experiments under various conditions. Testing apparatuses for direct shear, triaxial compression, and biaxial shear were mainly used and improved to reproduce the thermo-hydro-mechanical conditions of local bedrock, and it was confirmed that the shear behavior could vary. In order to design a high-level radioactive waste disposal site in Korea, the long-term behavior of rock discontinuities should be investigated in consideration of rock types, thermo-hydro-mechanical conditions, metamorphism, and restoration of shear resistance.

• Journal of the Mineralogical Society of Korea
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• v.17 no.1
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• pp.99-114
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• 2004

Mineralogical characteristics of fracture-filling minerals from deep borehole in the Yuseong area were studied for the radioactive waste disposal project. There are many fracture zones in the deep drill holes of the Yuseong granite, which was locally affected by the hydrothermal alteration. According to the results of hole rock analysis of drill core samples, $SiO_2$ contents are distinctly decreased, whereas $Al_2$$O_3$ and CaO contents and L.O.I. values are increased in the -90 m∼-130 m and -230 m∼-250 m zone, which is related to the formations of filling minerals. Fracture-filling minerals mainly consist of zeolite minerals (laumontite and heulandite), calcite, illite ($2M_1$ and 1Md polytypes), chlorite, epidote and kaolinite. The relative frequency of occurrence among the fracture-filling minerals is calcite zeolite mineral > illite > epidote chlorite kaolinite. Judging from the SEM observation and EPMA analysis, there is no systematic change in the texture and chemical composition of the fracture-filling minerals with depth. In the study area, low temperature hydrothermal alteration was overlapped with water-rock interactions for a long geological time through the fracture zone developed in the granite body. Therefore the further study on the origin and paragenesis of the fracture-filling minerals are required.

• Tunnel and Underground Space
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• v.21 no.6
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• pp.518-525
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• 2011

In this study, the natural ventilation pressure resulting from the large altitude difference which is a characteristic of high radioactive waste repository and the caloric value of the heat emitted by wastes was calculated and based on the results, natural ventilation quantities were calculated. A high radioactive waste repository can be considered as being operated through closed cycle thermodynamic processes similar to those of thermal engines. The heat produced by the heating of high radioactive wastes in the underground repository is added to the surrounding air, and the air goes up through the upcast vertical shaft due to the added heat while working on its surroundings. Part of the heat added by the work done by the air can be temporarily changed into mechanical energy to promote the air flow. Therefore, if a sustained and powerful heat source exists in the repository, the heat source will naturally enable continued cyclic flows of air. Based on this assumption, the quantity of natural ventilation made during the disposal of high radioactive wastes in a deep geological layer was mathematically calculated and based on the results, natural ventilation pressure of $74{\sim}183$Pa made by the stack effect was identified along with the resultant natural ventilation quantity of $92.5{\sim}147.7m^3/s$. The result of an analysis by CFD was $82{\sim}143m^3/s$ which was very similar to the results obtained by the mathematical method.

• 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.

• Tunnel and Underground Space
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• v.32 no.6
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• pp.530-548
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• 2022

The site selection process for deep geological disposal of high-level radioactive waste will be conducted in stages, and 103 evaluation parameters related to site selection have been proposed. In the field of rock mechanics and rock engineering, there are 33 evaluation parameters for intact rock, joint and rock mass, and they are applied in the basic and detailed investigation stages. In this report, uniaxial compressive strength, in-situ stress, joint distribution, and rock mass classification were selected as the main evaluation parameters, and among them, uniaxial compressive strength and in situ stress were selected as key evaluation parameters. Statistical techniques or regression analysis were performed for granite in Wonju and Chuncheon to evaluate the distribution range for the selected key evaluation parameters. The average of the uniaxial compressive strength in the Wonju area estimated through the posterior distribution is about 171 MPa, and about 123 MPa in the Chuncheon area. The maximum in situ stress acting in the Wonju area was less than 30 MPa and less than 40 MPa in the Chuncheon area. The direction of the maximum horizontal stress calculated by regression analysis was 101° in Wonju, and in the case of Chuncheon, it was 95°, respectiviely.