Disposal Of High-level Radioactive Waste Research Articles

Development of thermal-hydraulic coupling model for deep-geological disposal of high-level radioactive wastes

The buffer plays an important role in a engineered barrier system for the deep geological repository of high-level radioactive waste (HLW). High temperature may result in degradation of buffer performance. It is crucial to investigate the buffer temperature behavior so that its peak temperature is not higher than the criterion of 100 °C. In the present work, the thermal-hydraulic (T-H) coupling model is developed to simulate the histories of temperature and saturation in the disposal repository. This model is validated with the previous simulation works in the DECOVALEX 2015 Task B2 in order to simulate the geological repository of HLW with confidence. The whole repository of HLW in the present works can be reasonably simplified as a single hole with the symmetric boundary conditions on its peripheral sides. Effects of tunnel spacing and hole spacing of repository on the histories of buffer peak temperature are considered since the layout with various spacing is strongly related to the peak temperature of buffer. Based on the comparison of simulation results from the T-H coupling model and T-only model for the different layout of whole HLW repository, the maximum value of peak temperature in the buffer predicted by the T-H coupling model is lower than that from the T-only model. In other words, less area is needed for the selection of repository, rendering that the present T-H coupling model can assist in the design of HLW repository.

Changes of Temperature and Moisture Distribution over Time by Thermo-Hydro-Chemical (T-H-C)-Coupled Analysis in Buffer Material Focusing on Montmorillonite Content

Bentonite is used as a buffer material in engineered barriers for the geological disposal of high-level radioactive waste. The buffer material will be made of bentonite, a natural clay, mixed with silica sand. The buffer material is affected by decay heat from high-level radioactive waste, infiltration of groundwater, and swelling of the buffer material. The analysis of these factors requires coupled analysis of heat transfer, moisture transfer, and groundwater chemistry. The purpose of this study is to develop a model to evaluate bentonite types and silica sand content in a unified manner for thermo-hydro-chemical (T-H-C)-coupled analysis in buffer materials. We focused on the content of the clay mineral montmorillonite, which is the main component of bentonite, and developed a model to derive the moisture diffusion coefficient of liquid water and water vapor based on Philip and de Vries, and Kozeny–Carman. The evolutions of the temperature and moisture distribution in the buffer material were analyzed, and the validity of each distribution was confirmed by comparison with the measured data obtained from an in situ experiment at 350 m in depth at the Horonobe Underground Research Center, Hokkaido, Japan.

Performance and mechanism of siderite/zero-valent nickel composite material for removing Se(IV) from solution

79Se is a fission product in spent fuel with a half-life of up to 6.5 × 104 a, which is not controlled by solubility in aqueous solution and has strong mobility. These characteristics make it highly susceptible to migration in groundwater. Meanwhile, due to the negative charge of selenium ions, they are difficult to be fixed by buffer/backfill materials, engineering materials, geological materials and commonly used adsorption methods in near-field geochemical environments. Therefore, the fixation of 79Se has become a hot and difficult research topic both domestically and internationally. Therefore, this article aimed to simulate the removal of radioactive nuclide Se(IV) from high-level radioactive waste disposal repositories. On the basis of siderite (FeCO3), by adding zero-valent nickel (Ni0) and continuously optimizing sample preparation conditions, an economical, efficient and stable siderite/zero-valent nickel (FeCO3/Ni0) composite material was successfully prepared. Experimental studies on the removal of Se(IV) from solution were conducted under different conditions and the material was characterized using various modern methods from macro and micro perspectives, the performance and mechanism of FeCO3/Ni0 composite material for removing Se(IV) was explored deeply. The research results indicated that FeCO3/Ni0 had high potential as a candidate adsorbent for the removal of Se(IV) due to its easy synthesis, low-cost, high adsorption capacity and strong reduction.

Curing temperature’s effect on argillite-metakaolin based geopolymer grouts

This study concerns the project of high and intermediate level radioactive waste disposal (Cigéo project for Centre Industriel de stockage GEOlogique). Millions of tons of argillite will be excavated during the construction of the site. Therefore, the aim of this work is to reuse this argillite in the formulation of an injectable geopolymer grout and study its evolution in different storage conditions (T = 20 and 80 °C). To achieve this challenge, a metakaolin and calcined Callovo-Oxfordian (COx) argillite were used with potassium silicate solutions. The grouts were successfully obtained from both metakaolin and COx argillite in both temperatures 20 and 80 °C. The range of viscosity of several mixtures respected the required specifications (viscosity ≤5 Pa s and setting time superior to 6 h). The metakaolin-based geopolymers didn’t display any major changes no matter the storage temperature, except for the pore size, a coalescence phenomenon was observed with the increase of temperature. In contrast, for argillite or argillite-metakaolin based samples, a decrease of pore size as well as in pH values were observed with the increase of temperature. This behaviour was attributed to the dissolution of argillite’s impurities such as illite, providing more oligomers to allow the formation of different geopolymer networks.

Review of the Modelling of Corrosion Processes and Lifetime Prediction for HLW/SF Containers—Part 1: Process Models

The disposal of high-level radioactive waste (HLW) and spent nuclear fuel (SF) presents a unique challenge for the prediction of the long-term performance of corrodible structures since HLW/SF containers are expected, in some cases, to have lifetimes of one million years or longer. Various empirical and deterministic models have been developed over the past 45 years for making predictions of long-term corrosion behaviour, including models for uniform and localised corrosion, environmentally assisted cracking, microbiologically influenced corrosion, and radiation-induced corrosion. More recently, fracture-mechanics-based approaches have been developed to account for joint mechanical–corrosion degradation modes. Regardless of whether empirical or deterministic models are used, it is essential to be able to demonstrate a thorough mechanistic understanding of the corrosion processes involved. In addition to process models focused on specific corrosion mechanisms, there is also a need for performance-assessment models as part of the overall demonstration of the safety of a deep geological repository. Performance-assessment models are discussed in Part 2 of this review.

Public service motivation and not-in-my-backyard: The case of high-level radioactive waste disposal sites in Japan

This study investigated the association between public service motivation (PSM) and the not-in-my-backyard (NIMBY) reaction by the public that often blocks important projects. A conjoint experiment was conducted using the scenario of hypothetical plans for the construction of a high-level radioactive waste disposal site, one of the most noxious NIMBY facilities. The results showed that overall PSM (the full scale) had no association with the public’s willingness to accept such facilities. Still, we observed significant positive correlations for the PSM dimensions of attraction to public service and self-sacrifice, and a significant negative correlation with the PSM dimension of compassion. As expected, some dimensions of PSM strengthened (moderated) the positive effect of policy attributes on acceptance. Contrary to our prediction that PSM attenuates the relationship between distance and acceptance, overall PSM, commitment to public values, and compassion amplified the relationship. Together, these findings suggest the need to promote research considering the multifaceted nature of PSM in NIMBY problems.

Safety assessment in the disposal of high-level radioactive wastes (HLWs): a geochemical study of uranium complexes in deep groundwater in granites from Beishan, China

Safety assessment in the disposal of high-level radioactive wastes (HLWs): a geochemical study of uranium complexes in deep groundwater in granites from Beishan, China

Iodide retention characteristics on oxidized Cu coupon in saline and alkaline environments: Perspectives on high-level radioactive waste disposal

Radioactive iodine, particularly 129I, is one of the most hazardous radioisotopes found in deep geological repositories for high-level radioactive waste (HLW). This study investigated the iodide retention capacity of Cu coupons, simulating Cu canisters, under potential repository conditions. In saline environments (0.1 M NaCl + 0.01 M MgSO4), iodide was immobilized through CuI(s) formation on the oxidized surface of Cu coupons. This process was facilitated by the favorable affinity between Cu+ oxidized from the Cu coupons and I−. The iodide retention capacity of Cu coupons exhibited notable sensitivity to the pH and chemical composition of the solutions. Cu+ remained stably complexed with Cl− under saline conditions, leading to the CuI(s) formation. However, in distilled water, iodide retention decreased over time due to the transformation of CuI to CuO through Cu oxidation. In an alkaline solution (0.1 M KOH), iodide retention was negligible. These results indicate that radioactive iodide ions, released in the event of canister failure, can be effectively immobilized, and their migration to the ecosystem can be impeded through CuI(s) formation on oxidized Cu canisters in saline conditions. Consequently, Cu canisters can serve as an effective barrier against radioactive iodide migration and as a long-term isolation medium for HLW, particularly under saline conditions.

Modelling of failure and fracture development of the Callovo-Oxfordian claystone during an in-situ heating experiment associated with geological disposal of high-level radioactive waste

To ensure the safety of geological disposal of high-level radioactive waste, in-situ experiments have been carried out to examine the behavior of rocks in underground research laboratories (URLs). At the Meuse/Haute-Marne URL in France, the French National Radioactive Waste Management Agency (Andra) has been assessing the Callovo-Oxfordian claystone (COx) as potential host rock of geological disposal by subjecting the COx to in-situ heating mimicking exothermic radioactive waste. Results of the in-situ experiments are used to validate and bolster the numerical simulators for predicting the thermo-hydromechanically (THM) coupled behavior of the COx. The numerical simulators are, however, yet to be tested for predicting the failure and fracture development of the COx during heating, which is of paramount importance to the safety of the geological disposal. In this research, we modelled a recently carried out in-situ experiment at the Meuse/Haute-Marne URL using the TOUGH-FLAC simulator to predict the failure and fracture development of the COx during heating. The objectives are to examine the effects of (i) the weak bedding planes, (ii) the softening rate of matrix/weak plane strengths, and (iii) the stiffness anisotropy of the COx on the development of shear and tensile fractures during heating. Results show that considering failure along the weak planes enabled accurate predictions of fracture development. Also, fracture development intensified at a softening rate beyond a threshold level and the geometry of fractures was significantly affected by the stiffness anisotropy. These results will help boost the reliability of the safety and performance assessment of geological disposal in claystone.

Effects of temperature and confining pressure on the permeability of Beishan granite from high-level radioactive waste disposal repository

Understanding the variation of rock permeability with temperature and confining pressure is the foundation for projects such as high-level radioactive waste disposal, deep geothermal exploitation, and carbon dioxide storage. This paper presents an experimental study of the permeability evolution of thermally damaged Beishan granite under effective stress. The results show that temperature has a significant impact on the permeability of granite, and there is an obvious threshold temperature (400–500 °C), initial permeability and permeability change rate increased gradually with temperature. The permeability of Beishan granite shows non-linear decreasing trend with the increase of effective stress, ultimately tending towards stability. The microstructure of granite after thermal damage can be observed by scanning electron microscopy and used to assist in explaining the microscopic mechanism of permeability changes. A new granite permeability prediction model is proposed, which integrates permeability with temperature and effective stress, providing an important reference for the safety of high-level radioactive waste disposal.

Scenario development for safety assessment in deep geologic disposal of high-level radioactive waste and spent nuclear fuel: A review.

Radiation and radioactive substances result in the production of radioactive wastes which require safe management and disposal to avoid risks to human health and the environment. To ensure permanent safe disposal, the performance of a deep geological repository for radioactive waste is assessed against internationally agreed risk-based standards. Assessing postclosure safety of the future system's evolution includes screening of features, events, and processes (FEPs) relevant to the situation, their subsequent development into scenarios, and finally the development and execution of safety assessment (SA) models. Global FEP catalogs describe important natural and man-made repository system features and identify events and processes that may affect these features into the future. By combining FEPs, many of which are uncertain, different possible future system evolution scenarios are derived. Repository licensing should consider both the reference or "base" evolution as well as alternative futures that may lead to radiation release, pollution, or exposures. Scenarios are used to derive and consider both base and alternative evolutions, often through production of scenario-specific SA models and the recombination of their results into an assessment of the risk of harm. While the FEP-based scenario development process outlined here has evolved somewhat since its development in the 1980s, the fundamental ideas remain unchanged. A spectrum of common approaches is given here (e.g., bottom-up vs. top-down scenario development, probabilistic vs. bounding handling of uncertainty), related to how individual numerical models for possible futures are converted into a determination as to whether the system is safe (i.e., how aleatoric uncertainty and scenarios are integrated through bounding or Monte Carlo approaches).

Influence of radiation on borosilicate glass leaching behaviors

Vitrification is widely recognized as a promising method for the geological disposal of high-level radioactive waste (HLW) worldwide. To ensure the safe disposal of radioactive waste, the borosilicate glass that vitrifies HLW must exhibit exceptional water resistance to prevent the possibility of groundwater corrosion and subsequent radioactive leaks. Radiation might change the water resistance of borosilicate glass. A series of zirconium-containing borosilicate glass with an irradiation dose of 0.3 dpa were utilized to examine the radiation effect on glass-water interaction. Scanning electron microscopy (SEM), Time-of-Flight Secondary ion mass spectrometry (ToF-SIMS), X-ray photoelectron spectroscopy (XPS), Inductively Coupled Plasma Optical Emission spectroscopy (ICP-OES) and Fourier transform infrared spectroscopy (FTIR) were used to investigate the leaching behavior of the non- and irradiated samples. The depth profile of the leached samples implied the interdiffusion dominated glass-water interaction. The results from FTIR and ICP-OES indicated that, after irradiation, the initial leaching rate increased by threefold. Additionally, the impact of different zirconium contents on the water resistance of borosilicate glass was also presented.

STRUCTURAL FEATURES AND PROPERTIES OF MULTICOMPONENT SODIUM-CESIUM/RUBIDIUM ALUMINOBOROSILICATE MODEL MATRIX GLASSES OF BASIS COMPOSITION

Synthesis and study of the structure and various properties of Na-Cs and Na-Rb aluminoborosilicate model glasses with addition of the alkaline-earth cations have been carried out. Vibrational spectroscopy techniques have been used to establish the high degree of polymerization and noticeable structural similarity of all synthesized glasses at fixed ratio of modifying and network-forming cations. It has been shown that increasing fraction of large cations in the total content of modifying cations leads to changes in distribution of modifying cations between the silicate and borate structural units. The observed homogeneity with the similarity of anionic structure, at different content and various combinations of alkali and alkaline-earth cations, points at promising application of these glasses as a basis for synthesis of matrix materials for high-level radioactive waste disposal.

Modelling of the Corrosion-Induced Gas Impact on Hydraulic and Radionuclide Transport Properties of Geological Repository Barriers

The geological disposal of high-level radioactive waste is the final step in the nuclear fuel cycle. It is realized via isolating the high-level radioactive waste in the geological environment with an appropriate system of engineered barriers. Radionuclides-containing materials must be isolated from the biosphere until the radioactivity contained in them has diminished to a safe level. In the case of high-level radioactive waste, it could take hundreds of thousands of years. Within such a long timescale, a number of physical and chemical processes will take part in the geological repository. For the assessment of radionuclide migration from a geological repository, it is necessary to predict the repository’s behavior once placed in the host rock as well as the host-rock response to disturbances due to construction. In this study, the analysis of repository barriers (backfill, concrete, inner excavation disturbed zone (EDZ), outer EDZ, host rock) thermo–hydraulic–mechanical (THM) evolution was performed, and the scope of gas-induced desaturation was analyzed with COMSOL Multiphysics. The analysis was based on modelling of a two-phase flow of miscible fluid (water and H2) considering important phenomena such as gas dissolution and diffusion, advective–diffusive transport in the gaseous phase, and mechanical deformations due to thermal expansion of water and porous media. The importance of proper consideration of temperature-dependent thermodynamic properties of water and THM couplings in the analysis of near-field processes was also discussed. The modelling demonstrated that such activities as 50 years’ ventilation of the waste disposal tunnel in initially saturated porous media, and such processes as gas generation due to corrosion of waste package or heat load from the waste, also led to desaturation of barriers. H2 gas generation led to the desaturation in engineered barriers and in a part of the EDZ close to the gas generation place vanishing soon after finish of gas generation, while the host rock remained saturated during the gas generation phase (50–100,000 years). Radionuclide transport properties in porous media such as effective diffusivity are highly dependent on the water content in the barriers determined by their porosity and saturation.

Using multiple isotopes to determine groundwater source, age, and renewal rate in the Beishan preselected area for geological disposal of high-level radioactive waste in China

Understanding hydrogeological conditions is crucial for selecting and assessing the long-term safety performance of a high-level radioactive waste (HLW) disposal repository. Utilizing environmental isotopes as effective markers for analysing groundwater movement, this study investigates groundwater recharge sources, age, and renewal rates using multiple isotopes in China's potential HLW repository site, the Beishan area. The results indicated deep bedrock groundwater primarily derives from ancient precipitation infiltration under cold climatic conditions. A noteworthy distinction is that loose sedimentary groundwater exhibits higher tritium content (>10 TU) compared to bedrock groundwater (<3.2 TU). Groundwater within the recharge area, especially within gullies and piedmont slope deposits, is relatively youthful, with an age of less than 30 years and an annual renewal rate exceeding 5 %. In contrast, the shallow groundwater age in the intermountain basins and depressions of the discharge area generally exceeds 50 years, with an annual renewal rate often falling below 0.5 %. At the Beishan underground research laboratory site, deep groundwater at the disposal repository depth displays a corrected 14C age exceeding 8,000 years, indicating an extremely slow movement and alteration rate. As a result, the hydrogeological conditions in the Beishan area are expected to be relatively beneficial for ensuring the safety of HLW repository.

Prediction of Long-Term Geochemical Change in Bentonite Based on the Interpretative THMC Model of the FEBEX In Situ Test

Since nuclear energy is crucial in the decarbonization of the energy supply, one hurdle to remove is the handling of high-level radioactive waste (HLW). Disposal of HLW in a deep geological repository has long been deemed a viable permanent option. In the design of a deep geological repository, compacted bentonite is the most commonly proposed buffer material. Predicting the long-term chemical evolution in bentonite, which is important for the safety assessment of a repository, has been challenging because of the complex coupled processes. Models for large-scale tests and predictions based on such models have been some of the best practices for such purposes. An 18-year-long in situ test with two dismantling events provided a unique set of chemical data that allowed for studying chemical changes in bentonite. In this paper, we first developed coupled thermal, hydrological, mechanical, and chemical (THMC) models to interpret the geochemical data collected in the in situ test and then extended the THMC model to 200 years to make long-term prediction of the geochemical evolution of bentonite. The interpretive coupled THMC model shows that the geochemical profiles were strongly affected by THM processes such as evaporation/condensation, porosity change caused by swelling, permeability change, and the shape of concentration profiles for major cations were largely controlled by transport processes, but concentration levels were regulated by chemical reactions, and the profiles of some species such as pH, bicarbonate, and sulfate were dominated by these reactions. The long-term THMC model showed that heating prolongs the time that bentonite becomes fully saturated in the area close to the heater/canister; however, once the bentonite becomes fully saturated, high concentrations of ions in bentonite near the heater, which was observed in the field test, will disappear; illitization continues for 50 years but will not proceed further.

Speciation studies at the Illite - solution interface: Part 1 – Sorption of phosphate ions

Although phosphate ion (hereafter “P”) sorption is known to influence clay surface reactivity and trace metal behavior in the environment, there is still a lack of in situ spectroscopic evidence on the identity of phosphate sorption species formed at the interface between solution and clays with a 2:1 layered structure (2 silica tetrahedral: 1 octahedral layers), which are ubiquitous in soils and clay formations. We studied the surface speciation of phosphate ions by in situ spectroscopic monitoring along P sorption at the interface between a solution and a homoionic Na-illite, as a function of time or aqueous phosphate concentration (20–250 µM) leading to low/moderate P surface coverage. We also combined them with batch sorption experiments and electrophoretic mobility (EM) measurements. Macroscopic data indicated that the percentage of P sorption depends on pH, aqueous phosphate concentration, and the clay/solution ratio of the experiment, i.e., the coverage of the clay surface by P, with a maximal sorption capacity of the clay equal to ca. 6 µmol.g−1 at pH 4. Macroscopic and EM data further suggest that P sorption mechanisms onto illite confer negative charges on the surface and involve several sorption species and/or surface sites present on the clay edges. In situ FTIR spectra showed that, at acidic pH, three types of phosphate surface species exist at the illite-solution interface: an outer-sphere surface complex (OSSC) of phosphate (ΞSOH2+…H2PO4-, with S referring to a sorption site on the clay), an inner-sphere surface complex (ISSC) of phosphate, probably a monodentate binuclear surface complex ((ΞSO)2PO2) formed by surface ligand exchange with high-affinity aluminol sites and, to a lesser extent, low-affinity sites on the clay edges, and, in limited amounts, an Al-phosphate surface precipitate. Phosphate ion sorption occurs through the initial formation of the OSSC of phosphate, which dominates phosphate surface speciation at short reaction times and low P concentrations and transforms over time into the ISSC of phosphate. This study presents the first published in situ surface speciation of phosphate ions sorbed at the interface between a solution and an illite. The evidence of a strong sorption of phosphate ions at this clay surface provides new and sound knowledge to better understand the environmental fate of phosphate ions and trace metals, particularly in agricultural soil - groundwater continua or in clay host rocks considered for high-level radioactive waste disposal.

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.

Radiation effect on the corrosion of disposal canister materials

The effects of radiation on the corrosion of canister materials were investigated for the reliable disposal of high-level radioactive waste. The test specimens were gamma-irradiated at a very low dose rate of approximately 0.1 Gy/h for six and twelve months. The copper and cast iron species were less corroded when irradiated. It is hypothesized that gamma rays suppress the formation of lower-enthalpy species like metal oxides and activate reductive reactions. In contrast, it was difficult to evaluate the effect of radiation on the corrosion of titanium and stainless steel.

Evaluation of EDZ characterization technology for URL based on field investigation at BET

Detailed characterization of the Excavation Damage Zone (EDZ) is an important issue for the high-level radioactive waste (HLW) geological disposal engineering, which is closely related to the long-term safety of the HLW repository. With drill-blast experiment in Beishan Exploration Tunnel (BET, a platform for the field experiment and validation of potential technologies used for China’s HLW disposal engineering in crystalline rock), systematic field tests were carried out to evaluate the suitability of techniques in characterizing EDZ, and to study EDZ characteristic in crystalline tunnel. A series of investigation technologies were used to delineate EDZ and to examine the process of its formation, which includes: p-wave velocity measurement, ground penetrating radar (GPR), acoustic emission (AE) monitoring and borehole televiewer. P-wave velocity measurement along borehole reveal that wave velocity decreased significantly (more than 10%) in 30 cm to 50 cm from the excavation surface, and this zone is defined as EDZ. Clear images of borehole televiewer were obtained, which is helpful for EDZ delineation. GPR and borehole radar were also used for EDZ delineation. Strong reflection signals were detected in the range of 10 cm to 25 cm by GPR with high frequency antenna, indicating significant changes of physical properties of rock-mass in the zone, which is defined as Highly Damaged Zone (HDZ). Moreover, the formation process of EDZ induced by blast in crystalline rock tunnel was analyzed using AE monitoring. Damage process of surrounding rock was divided into blasting damage stage and progressive damage stage. AE events location show the induced cracking activities were clustering within 30 cm to 100 cm from excavation surface. With the experience from field investigation and result analysis, the feasibility and suitability of the techniques for EDZ characterization is evaluated. Finally a specialized EDZ characterization technique system is proposed for underground research laboratory construction in crystalline host rock in Beishan area.