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Highlights • The numerical modeling for a repository was validated using an analytical model. • Thermo-hydraulic behavior of a bentonite buffer in a repository was investigated. • The effect of design factors for a buffer was evaluated through sensitivity analysis. • A case study of sensitivity analysis was conducted according to a buffer property. Abstract A buffer in a geological repository minimizes groundwater inflow from the surrounding rock and protects the disposed high-level waste against any mechanical impact. As decay heat of spent fuel causes temperature variation in the buffer, which affects the mechanical performance of the system, an accurate estimation of the temperature variation is substantial. The peak temperature of the buffer, which is a critical design criterion of the repository, is affected by not only the thermal properties but also the hydraulic properties of the buffer, and these properties also affect each other. Thus, the effect of these factors should be properly evaluated during the system design. Hence, in this study, a sensitivity analysis of seven input factors was conducted using a numerical model based on a finite element method, and the model was verified using an analytical solution. In addition, a case study of the sensitivity analysis was conducted on two different property relationships of the buffer. The sensitivity indices of thermal conductivity, intrinsic permeability, and viscosity, which were determined as major factors, were 0.387, 0.215, and 0.163 in Case 1, and 0.510, 0.145, and 0.110 in Case 2, respectively, which means that these properties should be carefully considered during the design process of repositories. Furthermore, the case study shows that the relationship of the properties of a bentonite buffer has a significant effect on the design properties influencing the temperature variation. [ABSTRACT FROM AUTHOR]

The decay of actinides such as 238 Pu, results in recoil damage and helium production in spent nuclear fuels. The extent of the damage depends on storage time and spent fuel composition and has implications for the integrity of the fuels. Some advanced nuclear fuels intended for use in pressurized heavy water pressure tube reactors have high initial plutonium content and are anticipated to exhibit swelling and embrittlement, and to accumulate helium bubbles over storage times as short as hundreds of years. Calculations are performed to provide estimates of helium production and fuel swelling associated with alpha decay as a function of storage time. Significant differences are observed between predicted aging characteristics of natural uranium and the advanced fuels, including increased helium concentrations and accelerated fuel swelling in the latter. Implications of these observations for long term storage of advanced fuels are discussed. [ABSTRACT FROM AUTHOR]

• Comprehensive work on the DDML for the disposal of the HLW is examined. • Typical DDML algorithms and the main characteristics for the DGR are summarized. • Applications and development of the DDML for the disposal of HLW are analyzed. The application of the data-driven machine learning (DDML) for the disposal of the high-level nuclear waste (HLW) is of emerging interest in the recent years. This review aims to systematically elaborate, analyze, and summarize recent advances related to DDML in the area of disposal of the HLW. Firstly, a comprehensive work on the DDML for the disposal of the HLW is examined. Five DDML algorithms including the linear regression (LR), principle component analysis (PCA) and artificial neural network (ANN) are illustrated. Then, it summarizes the typical DDML algorithms and the main inputs/outputs for the deep geological repository (DGR). Furthermore, it is concluded that the hybrid DDML algorithms are efficient choices. Also, the DDML shows a great applicability for the simulation of the multiscale and multiphysics field. Lastly, the physical-informed DDML may enhance the performance of all algorithms. [ABSTRACT FROM AUTHOR]

Albedo neutron dosimeter utilizes the property of the human body to moderate incident high energy neutrons, which are scattered and reflected as thermal and low energy neutrons that can be measured by thermoluminescence detector (TLD) with high sensitivity. The current study is devoted to the assessment of possible utilization of the albedo neutron dosimeter as a personal monitoring device in a deep geological disposal system for high-level nuclear waste. When a nuclear waste package is transported to its final disposal position in a deep geological repository, such as in a horizontal emplacement drift in a rock salt mine, the question about the appropriate calibration factor of the albedo neutron dosimeter arises. Neutrons backscattered by the surrounding host rock layers might contribute significantly to the composition of the radiation field and in turn influence the response of the neutron detector. To account for different radiation field compositions, a field calibration technique providing field-specific calibration factors was applied in the current study. The field calibration of an albedo neutron dosimeter in a deep geological repository was performed with Monte Carlo method. The radiation field was simulated by a shielding cask loaded with a 252 Cf neutron source that is placed in a horizontal drift of a rock salt mine. Calibration factors were calculated at various distances to the shielding cask. In the investigated case, the neutron radiation field in a geological repository can be assigned to the calibration factor N1, which is recommended for reactors and accelerators with heavy shielding. [ABSTRACT FROM AUTHOR]

• Pu/Th spent fuel needs three to four times larger DGR compared with uranium fuel. • Equations to estimate minimum canister spacing as function of power was derived. • The optimal number of channels per canister for Pu/Th is 4 while it is 12 for UOX. • No criticality hazard exists for storage of Pu/Th even if fresh fuel was considered. The Canadian conceptual design of the Super-Critical-Water Reactor (SCWR), known as Pressure-Tube –SCWR (PT-SCWR) is a GEN-IV design that employs the Pu/Th direct fuel cycle. The spent Pu/Th fuel is expected to be stored into Deep Geological Repository (DGR). In this study, a thermal analysis for a repository with Pu/Th spent fuel was performed using the computer code CODE_BRIGHT. The maximum temperature of the bentonite was obtained, and the minimum canister spacing was calculated for different cooling times and different numbers of channels per canister. The size of the repository needed for Pu/Th fuel was found to be three to four times larger than that for uranium for the same electrical energy output. This increases the cost of the long-term management of the spent fuel. The nuclear criticality safety aspect of storing Pu/Th spent fuel in the DGR was evaluated as well using MCNP. The effective multiplication factor, for the most reactive situations was 0.858, which is less than the acceptance criterion of 0.95. [ABSTRACT FROM AUTHOR]

[Display omitted] • Nuclear power is an ideal option for sustainable energy sources from U-235 fission. • A 1,000 MW e reactor generates about 25–30 m3 of PUNF annually. • Deep geological repository is the most ideal disposal technology for PUNF waste. • Th-232 is an ideal option to replace U-238 as a nuclear fuel. • Vitrification is prone to the risk of nuclear proliferation. Nuclear power is an ideal option for sustainable energy sources from U-235 fission. However, this energy generates long-term radioactive waste such as partially used nuclear fuel (PUNF) during electricity production. This work reviews various technologies to provide viable, sustainable, and long-term solutions for the PUNF storage. They include vitrification, partitioning and transmutation (P&T), pyro-processing, and deep geological repository (DGR). Their benefits and drawbacks are evaluated and compared based on previous studies. How to deal with the public perception of DGR and its impacts on the future of nuclear energy and the business opportunities of nuclear storage technology in the global market are discussed. A perspective of recycling nuclear waste into usable fuel is also elaborated. Our literature survey of 160 published articles (1981–2021) showed that DGR is the most ideal solution for long-term storage of the PUNF, as it provides an ultimate destination in a deep underground that permanently isolates the waste from inhabitants and the environment. Although storing PUNF in a DGR maybe convenient and economical in the short-term, the waste must be stored in a retrievable form so that it can be recycled as a fuel. In the long-term, a complete recycling of used nuclear fuel is the best option. As technological solutions and sound radioactive waste management policy are important for the safe storage of PUNF, stakeholders in the nuclear industry should portray long-term radioactive waste management through viable, feasible, and permanent solutions to waste storage for the sake of public safety and the environment. [ABSTRACT FROM AUTHOR]

• The influence of thermal and chemical gradient saturated on bentonite is presented. • A lab-scale set up was fabricated which would create a thermal and chemical gradient. • The degree of saturation decreased towards heating end from the hydration end. • Reduction in cation exchange capacity was noted at various levels of samples. • The micro-structural investigations revealed the mineralogical changes. Compacted bentonite is being used as a buffer material for the safe dumping of high-level radioactive wastes (HLW) in the deep geological repository (DGR) due to its favorable physicochemical properties. However, the temperature variation, probably because of continuous heat emitting canister (initially −150 °C to 250 °C) and geothermal gradient, affect its performance. There would be bentonite iron interface reactions due to the corrosion of steel canister, which might cause reactive transport through the compacted bentonite buffer. Therefore, it is essential to know the influence of a combination of thermal as well as a chemical gradient on the reactive transport through the compacted bentonite buffer in terms of its physicochemical, thermomechanical, mineralogical stability. The purpose of this particular study is to establish the field conditions in the lab by fabricating a small scale set up which would create a thermal gradient on the compacted Barmer bentonite (B 1 and B 2) samples (compacted at 1.5 Mg/m3, 1.75 Mg/m3 and 2 Mg/m3 densities) by heating its one side at 60 °C, 110 °C, 150 °C and 200 °C temperatures for 28 days each and the system was hydrated by distilled water from the other side and was kept in touch with the metallic iron powder. It was observed that the degree of saturation of both bentonites decreased with an increase in temperature as well as density. However, the cation exchange capacity at lower density reduced more than that of higher density. The microstructural observation revealed the presence of impurities between the lamella of bentonite after 28 days that might be hematite. [ABSTRACT FROM AUTHOR]

• Thermally stable iron phosphate glass for immobilization of spent nuclear fuel was developed. • Iron phosphate glass can contain up to 15 mol% of CeO 2 , as a surrogate for UO 2. • Iron phosphate glass has good chemical durability and nuclear criticality safety. • Repository space can be saved 11 times more than direct disposal by vitrification. Iron phosphate glasses with low melting temperatures of ∼1300 °C were developed to immobilize spent nuclear fuels (SNFs). The glasses have densities of ∼3.15 g/cm3 and glass transition temperatures of ∼540 °C that are high enough to endure the temperatures in geological repositories. The waste loading of CeO 2 in the glass was ∼21.92 wt%, which corresponds to ∼30.38 wt% of UO 2 at the same molar ratio. Normalized elemental releases from the product consistency test were well below the regulated limit of 2 g/m2. Effective neutron multiplication factor k eff was 0.755, which is below nuclear criticality safety regulation (k eff = 0.95). Vitrified SNF occupies one-eleventh of the repository space compared to direct disposal of the same amount of UO 2 , excluding left-over low-level wastes. [ABSTRACT FROM AUTHOR]