Geological Disposal System Research Articles

Swelling Stress of Bentonite: Thermodynamics of Interlayer Water in K-Montmorillonite in Consideration of Alteration

The buffer material that makes up the geological disposal system of high-level waste swells by contact with groundwater and seals space with rock mass and fractures in rock mass. The buffer material has a function of mechanical buffer with rock pressure, and swelling stress is important in this case. The alteration of bentonite may occur due to the initial replacement of cations (Na+ ions) in the interlayer with K+ ions upon contact with groundwater, but there are no studies on the swelling stress of K-bentonite. In this study, the author prepared K-montmorillonite samples and obtained thermodynamic data on interlayer water as a function of water content using a relative humidity method. The swelling stress was analyzed based on a thermodynamic model developed in earlier studies and compared with measured data. The activity and the relative partial molar Gibbs free energy of porewater decreased with decreasing water content in the region, below approximately 15%. This behavior significantly differs from that of other ions, such as Na. The swelling stress calculated based on the thermodynamic model and date occurred in the region of high density of 1.9 Mg/m3 with montmorillonite partial density. It was indicated for the first time that K-bentonite scarcely swells under realistic design conditions.

Dynamic behavior of sorption of europium onto biotite flakes under the condition of saline groundwater

This study discussed the interaction between Eu and biotite in flakes under the coexistence of Na ions, considering saline groundwater. The sorption experiments were conducted by mixing an Eu solution with Na or K ions and biotite flakes at pH 5. Resultantly, the sorption of Eu gradually increased because of the diffusion into the layered structure of biotite flakes. Thereafter, the sorption amount of Eu with Na ions was less than that without Na ions, and Eu was hardly sorbed with K ions. Additionally, the apparent diffusion coefficients were estimated to be in the order of 10−13 m2/s based on a two-dimensional diffusion model. These values were lower than the effective diffusion coefficients in plutonic rock, which are employed to assess the performance of geological disposal systems. This indicates that the sorption of Eu in plutonic rock is restricted by the diffusion process in biotite flakes.Graphical abstract

Glass/steel/clay interactions in a simulated radioactive waste geological disposal system

Deep geological storage is the accepted solution for the final disposal of high-level radioactive waste therefore, it is necessary to study the host rock of the planned Hungarian waste repository and the materials involved in the engineered barriers. The main goal was to understand the characteristics and stability of the glass/steel/claystone system, from the structural properties of the vitrified waste (borosilicate glasses) to the clay response in the repository. Repository conditions were applied during the experiments to understand the chemical evolution of the system. A triplicate setup was kept at 80 °C for 3, 7 and 12 months and post-mortem characterization was performed. No alteration products were observed with scanning electron microscopy energy dispersive X-ray spectroscopy measurements on the surface of the glass and Fe or in the clay after the end of the experimental period. Based on the elemental analysis of the liquid phase, the released amount of B, K, Si and Na increased, while that of Ca and Mg decreased compared to the baseline. The concentrations of Cl− and SO42− did not change significantly. Ca- and Mg-silicate precipitation was observed by X-ray photoelectron spectroscopy at the surface range of the borosilicate glasses because of the synthetic porewater treatment.

Time Dependent Evolution of Heat and Stress Field in Granite–clay System Representing Geological Disposal Scheme for High Level Radioactive Wastes

Abstract A reference Geological Disposal System (GDS) for high level long lived heat emitting radioactive waste with dimension 20 x 10 x 6.25 m has been evaluated for time dependent heat field distribution, stress build up and possibility of micro-fracturing. The GDS comprises of three cylindrical heat sources (0.30 m diameter and 2 m length) mimicking radioactive waste loaded canisters emplaced within Jalore granite (JG) with barrier layers of Barmer bentonite (BB) clays, from Rajasthan. The methodology adopted in the study includes thermo- mechanical characterization of the granites and clays, followed by finite difference method (FDM) based numerical analysis for time dependent buildup of thermal and mechanical stresses, evolution of temperature field and micro fracturing phenomena within the studied GDS. The study reveals that three heat sources (waste filled SS canisters) of 500 watt intensity with spacing of 2.5 m and clay granite (CG) ratio of 60:40 and 70:30 produces maximum temperature of 92.42 and 94.69 °C and total stress (thermal and mechanical) of 24.77 and 24.97 MPa respectively within the system. 3.32 and 3.31 mm maximum displacement observed in the GDS for 60:40 and 70:30 CG ratio respectively which is within design limit. Analysis further establishes that clay admixture having bentonite clay and granite ratio in the range of 70:30 and 60:40 is capable of smooth dissipation of heat thorough them with resultant maximum temperature of < 100 °C throughout the entire duration of the time periods analysed. Based on these results it is concluded that JG and BB clay in 60:40 and 70:30 ratio with waste canister spacing of 2.5 m have appropriate characteristics suitable for hosting geological disposal system for heat emitting high level waste canisters. The study marks first analysis of multi-canister geological disposal system in Indian context.

Characterization of Compacted Ca- and Na-Bentonite with Copper Corrosion Products in the KAERI Underground Research Tunnel

In a deep geological disposal system, bentonite buffer material is an important barrier used to protect the disposal canister from the inflow of groundwater and prevent the outflow of radionuclides. This study aimed to characterize the mineralogical and chemical reactions of bentonite caused by copper corrosion of the canister in a radioactive waste repository. We investigated the d-spacings of montmorillonite in Gyeongju bentonite (Ca-type, KJ-I) under groundwater-saturated conditions over 10 years and compared their characteristics with those of Wyoming bentonite (Na-type, MX-80) in the Korea Atomic Energy Research Institute Underground Research Tunnel. Mineralogical investigations using X-ray diffraction and focused ion beam energy-dispersive spectroscopy indicated that no transformation of smectite or neo-formed clay phases occurred. In the Ca-type bentonite (KJ-I), the swelling was observed when it was in contact with rolled plate (RP) and cold-spray-coated (CSC) copper, with d-spacing expansions of 2.9% and 3.8%, respectively. In contrast, the Na-type bentonite (MX-80) showed d-spacing expansions of 17.6% and 19.6% when it was in contact with the RP and CSC Cu, respectively. The Cu concentration and distribution indicated that the corrosion products dissolved and then diffused into the surrounding bentonite, with maximum penetration depths of 2.0 and 0.5 mm over 10 years, respectively.

KAERI underground research laboratory: Overview of in-situ experiments

Social concerns regarding the safety of high-level radioactive waste have increased with growing public awareness of environmental issues and nuclear power. The performance assessment of deep geological disposal systems is crucial to reduce the uncertainties associated with high-level radioactive waste disposal and enhance the overall public confidence in nuclear safety. Accordingly, the Korea Atomic Energy Research Institute (KAERI) has undertaken various studies on the development of a deep geological disposal system for high-level waste and disposal safety evaluation. The KAERI Underground Research Tunnel (KURT), South Korea's only underground research laboratory dedicated to radioactive waste disposal, was constructed in 2006 and expanded in 2015. Since its construction, numerous in-situ experiments have been conducted and are currently underway in the KURT. The KURT plays a significant role in assessing the feasibility, safety, stability and appropriateness of a deep geological disposal system in South Korea and also provides an opportunity to revitalize industrial-academic-scientific cooperation between related institutions. This report summarizes two key in-situ experiments and international joint research conducted between 2007 and 2017 to assess the performance of the engineered and natural barriers of the KURT. The research experiences from the in-situ tests conducted at the KURT will provide crucial information on the safety and feasibility validation of the deep geological disposal system and will be an important contributor to the success of the Korean high-level radioactive waste disposal program in the future.

Comparison of Numerical Analysis Methods of APro for the Total System Performance Assessment of a Geological Disposal System

Comparison of Numerical Analysis Methods of APro for the Total System Performance Assessment of a Geological Disposal System

A numerical study on THM coupled behavior in the high-level radioactive waste disposal system

It is essential to securely isolate high-level radioactive waste from the biosphere, and geologic disposal of it at a deep underground repository is considered the most effective method. Therefore, it is crucial to research the complex thermo-hydro-mechanical-chemical (THMC) coupled behavior in geological disposal systems parallel with numerical simulation. DECOVALEX is an international cooperating project to efficiently develop numerical methods and models and validate through test results for predicting the THMC interactions in the disposal systems. In Task C of DECOVALEX-2023, the modeling teams focus on understanding pore pressure development and THM interactions in the host rock and buffer material during the FE experiment, a 1:1 scale in-situ heater test based on Nagra’s Reference repository design. We used OGS-FLAC3D for the numerical simulation, combining OpenGeoSys for TH simulation and FLAC3D for M simulation. At the first phase of the task, a simple two-dimensional benchmark problem was defined to set up the numerical model. THM coupled processes in the bentonite were simulated with a two-phase flow system, and we investigated the temperature and pressure variations on the given monitoring position. Vaporization induced by a temperature increase in the bentonite was observed at the heater’s vicinity, and flow occurred by capillarity, and pressure difference was well simulated. Additionally, the flow process was dominant in the near field of engineering barrier, while we observed thermal pressurization in the far-field area. We plan to apply the developed model to a full-scale three-dimensional numerical simulation for the next phase.

Sorption Studies of Cesium on Kunipia-F Bentonite and Silicon Dioxide

Adsorption is one of the effective methods used for the removal of contaminants from radioactive wastewater. In this research, removal efficiency of Cs from the prepared low-concentration Cs solution (100 μg/L) was studied using two different adsorbents namely high purity Kunipia-F bentonite and Silicon dioxide (SiO2). After performing the adsorption tests, the Cs ions in liquid and solid phases were analyzed using the inductively coupled plasma mass spectrometry (ICP-MS) technique. The adsorption capacities of both materials were then compared. It was found that the Cs adsorption capacity of bentonite was higher than that of SiO2. The distribution coefficient of Cs in the bentonite was 1.0 L-sol/g-solid, which was 100 times greater than that of SiO2. The highest sorption of Cs on bentonite could be attributed to its smaller particle size (0.362 μm), compared to the SiO2 particle size (0.625 μm). Bentonite is natural clay and is generally used as a buffer material in geological disposal systems for high-level nuclear waste. The results of this research indicate that bentonite could be an effective absorbent for decontaminating low concentration Cs in radioactive wastewater.

A new methodology for thermal analysis of geological disposal of spent nuclear fuel using integrated simulations of gamma heating and finite element modeling

A new methodology is illustrated, where the evolution of temperature in a geological disposal system for spent nuclear fuel is estimated by integrated calculations of a spatially distributed gamma heating source with conventional finite element thermal transport modeling. A case with one canister loaded with fuel assemblies with a cooling time of 30 years in a KBS-3 type repository illustrates the methodology. For this particular case, the effect of including distributed gamma heating rate in the modeling has a small impact on the temperature distribution compared to the conventional case of heat generated locally in the canister, resulting in a small decrease of the maximum temperature in the canister. A large proportion of gamma heating occurs inside the outer boundary of the copper canister for this case. Other potential consequences of radiation escaping the canister are discussed.

An analytical model for 3D consolidation and creep process of layered fractional viscoelastic soils considering temperature effect

The behavior of long-term deformation of saturated soil under external loads has been one of the research focuses in geological engineering and geotechnical engineering. The temperature change, especially in the fields of geological radioactive waste disposal and geothermal system, usually significantly influence the consolidation and creep process of soils. This paper presents an analytical three-dimensional (3D) model for investigating the consolidation and creep behavior of layered soils considering temperature effect. Based on the analytical layer element elastic solution, fractional calculus theory and elastic-viscoelastic correspondence principle, an analytical 3D model is built by introducing the coefficients of thermal expansion and viscosity varying with temperature into the fractional Merchant model (FMM). Verifications against published analytical and numerical results are provided, followed by extensive parametric studies to investigate the effects of type of viscoelastic model, temperature change, fractional order and viscosity of FMM, layered characteristic of soils on consolidation and creep process. It’s evident that the present model predicts long-term behavior of layered viscoelastic soils well and can be used to evaluate the effect of temperature change on consolidation and creep process in geological engineering and geotechnical engineering.

A modularized numerical framework for the process-based total system performance assessment of geological disposal systems

This study developed a safety assessment tool for geological disposal systems called APro, a systemically integrated modeling system based on modularizing and coupling the processes which need to be considered in a geological disposal system. Thermal, hydraulic, chemical, canister failure, radionuclide release and transport processes were considered in the current version of APro. Each of the unit processes in APro consists of a single Default Module, and several Alternative Modules which can increase the flexibility of the model. As an initial stage of developing the modularization concept and modeling interface, the Default Modules of each unit process were described, with one Alternative Module of chemical process. The computation part of APro is mainly a MATLAB workspace controlling COMSOL and PHREEQC, which are coupled by an operator splitting scheme. The APro model domain is a stylized geological disposal system employing the Swedish disposal concept (KBS-3 type), but the repository layout can be freely adjusted. In order to show the applicability of APro to the total system performance assessment of geological disposal system, some sample simulations were conducted. From the results, it was confirmed that coupling of the thermal and hydraulic processes and coupling of the canister failure and the radionuclide release processes were well reflected in APro. In addition, the technical connectivity between COMSOL and PHREEQC was also confirmed.

Structural and spectroscopic studies of spontaneously formed crystalline Eu(iii)-aliphatic dicarboxylates at room temperature.

Complexation of actinides and lanthanides with carboxylic organic ligands is a critical issue affecting radionuclide migration from deep geological disposal systems of spent nuclear fuel. A series of Eu(iii)–aliphatic dicarboxylate compounds, as chemical analogs of radioactive Am(iii) species, Eu2(Ox)3(H2O)6, Eu2(Mal)3(H2O)6, and Eu2(Suc)3(H2O)2, were synthesized and characterized using X-ray crystallography and time-resolved laser fluorescence spectroscopy to examine the ligand-dependent binding modes and the corresponding changes in spectroscopic properties. Powder X-ray crystallography results confirmed that all of the compounds presented a crystalline polymer structure with a trigonal prism square-face tricapped polyhedron geometry centered on Eu(iii) in a nine-coordinate environment involving nine oxygen atoms. This study captures the transition of the coordination modes of aliphatic dicarboxylate ligands from side-on to end-on binding as the carbon chain length increases. This transition is illustrated in malonate bindings involving a combination of side-on and end-on modes. Strongly enhanced luminescence, especially for the hypersensitive peak, indicates a low site symmetry in the formation of solid compounds. The number of remaining bound water molecules was estimated from the resultant increased luminescence lifetimes, which were in good agreement with crystal structures. The excitation–emission matrix spectra of these crystalline polymers suggest that Ox ligands promote the sensitized luminescence of Eu(iii), especially in the UV region. In the case of Mal and Suc ligands, charge transfer occurs in the opposite direction from Eu(iii) to the ligands under UV excitation, resulting in weaker luminescence.

Uranium (VI) sorption on illite under varying carbonate concentrations: Batch experiments, modeling, and cryogenic time-resolved laser fluorescence spectroscopy study

Dissolved inorganic carbon (DIC) present in groundwaters can affect both the aqueous and surface species of hexavalent uranium (U(VI)) owing to its strong complexation ability with U(VI). However, the effect of DIC on U(VI) sorption on illite, which is a critical component of argillaceous rocks, is still unclear. In this study, we investigated the sorption of U(VI) on conditioned illite du Puy with varying DIC concentrations (up to 250 mM DIC) as a function of pH through batch sorption experiments, surface complexation modeling, and cryogenic time-resolved laser fluorescence spectroscopy (TRLFS). The distribution coefficients of U(VI) decreased with an increase in the DIC concentration. At relatively high DIC concentrations with respect to experimentally investigated range (≥100 mM DIC), no U(VI) sorption was observed. The U(VI) sorption on illite was modeled using the two-site protolysis non-electrostatic surface complexation and cation exchange model. Two ternary surface complexation reactions with carbonate were needed to depict the experimental sorption data in addition to binary and ternary hydroxo surface complexation reactions used to describe the U(VI) sorption to illite without carbonate. Cryogenic TRLFS data revealed that U(VI) did sorb to illite at high DIC concentrations (up to 10 mM DIC). The spectra were unchanged with varying DIC concentrations (2.0, 5.0, and 10 mM DIC) at pH ∼8.5, indicating that the surface speciation of U(VI) remained the same. The decay curves were biexponential, which further indicates that at least two surface species were responsible for the sorption. These results could improve the performance assessment of a deep geological disposal system in a sedimentary host rock.

Twenty years of research on spent nuclear fuel in the context of final disposal at KIT-INE in multinational collaborative projects

Abstract. Disposal of spent nuclear fuel (SNF) in deep geological repositories is considered a preferential option for the management of such wastes in many countries with nuclear power plants. With the aim to permanently and safely isolate the radionuclide inventory from the biosphere for a sufficient time, a multibarrier system consisting of technical, geotechnical and geological barriers is interposed between the emplaced waste and the environment. In safety assessments for deep underground repositories, access of water, followed by failure of canisters and finally loss of the cladding integrity is considered in the long-term. Hence, evaluating the performance of SNF in deep geological disposal systems requires process understanding of SNF dissolution and rates as well as quantification of radionuclides release from SNF under reducing conditions of a breached container. In order to derive a radionuclide source term, the SNF dissolution and alteration processes can be assigned to two steps: (i) instantaneous release of radionuclides upon cladding failure from gap and grain boundaries and (ii) a long-term release that results from dissolution of the fuel grains itself (Ewing, 2015). In this context, research at KIT-INE has focused for more than 20 years on the behavior of SNF (irradiated UO2 and MOX fuels) under geochemical conditions (pH, redox and ionic strength) representative of various repository concepts, including the interaction of SNF with backfill material, such as bentonite as well as the influence of iron corrosion products, e.g. magnetite and radiolytic reactions on SNF dissolution mechanisms. Since 2001, KIT-INE has contributed with experimental and theoretical studies on the behavior of SNF under repository relevant conditions to six Euratom projects viz SFS (2001–2004), NF-PRO (2004–2006), MICADO (2006–2009), RECOSY (2007–2011), FIRST-Nuclides (2012–2014) and DISCO (2016–2021). Moreover, since 2007, overall 4 consecutive projects for the Belgian waste management organization, ONDRAF-NIRAS, were performed on the behavior of SNF under conditions representative of the Belgian “Supercontainer” concept. In this contribution, we summarize major achievements of theses research projects to understand and quantify the radionuclide release from dissolving SNF under repository conditions. In particular, the dependence of radionuclide release on the chemical composition of the aqueous and gaseous phase in the proximity of repositories in different types of host rock is discussed.

Numerical Study on the Effect of Enhanced Buffer Materials in a High-Level Radioactive Waste Repository

In deep geological disposal system designs, it is important to minimize the installation area for cost effectiveness while satisfying the thermal requirements of the systems. An effective method to reduce the installation area for the systems is to employ an enhanced buffer material, as this can decrease the spacing between the disposal tunnels and deposition holes. Therefore, this study aims to evaluate the effect of an enhanced buffer material on the thermal behavior of the systems and their spacing. First, the discrete element method (DEM) was adopted to validate the thermal conductivity of the enhanced buffer material used, which was a mixture of bentonite and graphite. Then, a 3D finite element method (FEM) was conducted to analyze the proper disposal tunnel and hole spacing considering three cases with thermal conductivities values of the buffer as 0.8 W/(m K), 1.0 W/(m K), and 1.2 W/(m K). The results showed that the disposal tunnel and hole spacing could be reduced to 30 m and 6 m, respectively, when the temperature of the buffer surface facing the canister was 100 °C with a thermal conductivity value of approximately 1.2 W/(m K) or if more than 3% of graphite is added.

Temperature effect on the thermal and hydraulic conductivity of Korean bentonite buffer material

High-level radioactive waste (HLW) disposal is considered to constitute a disposal system within a deep rock mass using deep geological repository. A deep geological disposal system has an engineered barrier system (EBS) consisting of canisters, buffer material, and backfill material. Among these items, the buffer material protects a canister from groundwater inflow and prevents radionuclide outflow. It is an also very important factor for evaluating the stability of a disposal system in which heat is propagated from the canisters. The aim of this study was to evaluate the thermal and hydraulic properties of Gyeongju bentonite, a buffer material from Korea. The thermal conductivity and hydraulic conductivity of Gyeongju bentonite were measured according to different degrees of saturation and to dry density. The whole process is based on the temperature change induced in the disposal environment. The thermal conductivity increased as temperature increased, and as did the temperature effect with high initial degree of saturation. Additionally, the hydraulic conductivity also increased as temperature did, and decreased with high dry density. After the process of heating and cooling, the thermal and hydraulic conductivity of the bentonite presented irreversible changes.

Surface complexation of Ca and competitive sorption of divalent cations on montmorillonite under alkaline conditions

In the geological disposal system of high-level radioactive and transuranic wastes, an increase in Ca concentration (arising from the alteration of cementitious materials) could affect the retention of other radionuclides due to competitive sorption. Batch sorption experiments were performed to investigate the sorption behavior of Ca and its competition with other divalent cations (Sr and Ni) on the edge sites of Kunipia F montmorillonite under alkaline conditions. Ca sorption increased with pH (when pH > 8) and it was more pronounced under high ionic strength conditions, indicating that Ca formed a surface complexation with the edge sites. The sorption behavior of Sr was similar to that of Ca. Meanwhile, the Ni sorption increased with pH, when this was >6 (for lower pH values than in the case of Ca and Sr), Ni tends to form surface complexes more readily than Ca and Sr. The results of a fitting analysis conducted based on the 2-site protolysis non-electrostatic surface complexation and cation exchange (2SPNE SC/CE) model indicate that Ca and Sr sorb onto the ≡SW2OH sites and Ni sorbs onto the ≡SSOH sites. Additionally, the sorption of the CaOH+ and SrOH+ ionic pair species by cation exchange was implied at pH ~ 12. Competitive sorption experiments were also carried out to evaluate the effect of Ca on the sorption of Sr and Ni. Sr sorption decreased with Ca concentration in the alkaline pH region, whereas Ni sorption was not affected by Ca concentration. Overall, these results indicate that Ca and Sr sorb onto the same sites, while Ca and Ni sorb onto different sites; additionally, the competitive sorption seemed to depend on chemical similarities (e.g., hydrolysis behavior). The sorption model parameters obtained from the single-element batch sorption experiments successfully reproduced the results of the competitive sorption experiments.

Sorption and surface complexation modeling of 137Cs on Fe(II)-montmorillonite clay mineral relevant to nuclear waste disposal

Abstract Bentonite, a natural clay, is a promising candidate to be used as a buffer/backfill material in the geological disposal systems for management of high level waste (HLW). Fe released due to corrosion of canister/overpack may result in alteration of Na-montmorillonite (Na-Mt), present in bentonite, to Fe-montmorillonite (Fe-Mt) in a span of several years after the emplacement of vitrified HLW in deep geological repositories. For realistic performance assessment, it is essential to understand the sorption behavior of altered Mt with regard to the radionuclides present in HLW. Cs is one of the high-yield (137Cs t1/2 = 30.1 y, 6%) and long-lived (135Cs t1/2 = 2 × 106 y) fission products in spent fuels. The objective of present study is to understand the effect of various parameters, viz., time (0–48 h), pH (3.0–9.0), ionic strength (0.001–1 M) [Cs(I)] (10−10–10−3 M) and Fe dissolution on sorption behavior of Cs(I) on Fe(II)-Mt through batch sorption experiments. Fe(II)-Mt was synthesized by reducing Fe(III)-Mt using ascorbic acid as reducing agent in N2 atmosphere. The near-constancy in Cs(I) sorption on Fe(II)-Mt with pH (≥4), and decrease with increasing ionic strength, illustrate the ion exchange as dominant mode of Cs(I) sorption. Further, Cs(I) sorption isotherm on Fe(II)-Mt is found to be linear. The estimation of dissolved iron in the supernatant of Fe(II)-Mt suspensions demonstrated that dissolved Fe decreased with increase in pH and increased with increase in ionic strength. Moreover, the Fe2+/Fetotal ratio determined in all experiments was close to unity, thereby depicting that Fe(II) did not oxidize to Fe(III), except when suspension pH was ≥ 5.5. For comparison, Cs(I) sorption was also studied on Na(I)-Mt and compared with that on Fe(III)-Mt. Surface complexation modeling of Cs(I) sorption on the three clay minerals, viz., Na-Mt, Fe(II)-Mt and Fe(III)-Mt, has been successfully carried out.

Analyses on Thermal Stability and Structural Integrity of the Improved Disposal Systems for Spent Nuclear Fuels in Korea

With respect to spent nuclear fuels, disposal containers and bentonite buffer blocks in deep geological disposal systems are the primary engineered barrier elements that are required to isolate radioactive toxicity for a long period of time and delay the leakage of radio nuclides such that they do not affect human and natural environments. Therefore, the thermal stability of the bentonite buffer and structural integrity of the disposal container are essential factors for maintaining the safety of a deep geological disposal system. The most important requirement in the design of such a system involves ensuring that the temperature of the buffer does not exceed 100℃ because of the decay heat emitted from high-level wastes loaded in the disposal container. In addition, the disposal containers should maintain structural integrity under loads, such as hydraulic pressure, at an underground depth of 500 m and swelling pressure of the bentonite buffer. In this study, we analyzed the thermal stability and structural integrity in a deep geological disposal environment of the improved deep geological disposal systems for domestic light-water and heavy-water reactor types of spent nuclear fuels, which were considered to be subject to direct disposal. The results of the thermal stability and structural integrity assessments indicated that the improved disposal systems for each type of spent nuclear fuel satisfied the temperature limit requirement (< 100℃) of the disposal system, and the disposal containers were observed to maintain their integrity with a safety ratio of 2.0 or higher in the environment of deep disposal.