High-level Waste Research Articles

A CSS surface based THM bounding constitutive model for volumetric behavior of bentonite

Development of the constitutive model for bentonite under coupled thermo-hydro-mechanical (THM) condition is of great significance for the construction and safety assessment of deep geological disposal repositories for high-level radioactive waste (HLW). In this work, a new temperature-suction-mean net stress (T-s-p) space with the conception of critical saturated state (CSS) surface was defined to represent the actual stress state of bentonite under coupled THM condition. Then, based on the CSS surface, a THM constitutive model was proposed for describing the volumetric behavior of compacted bentonite. Under the THM model framework, two bounding surfaces were proposed to describe the elastoplastic volume changes induced by mechanical response of skeleton and hydration of montmorillonite, respectively. The model responses upon some typical THM paths were simulated and discussed to reveal the performance of the proposed constitutive model for bentonite. Finally, the proposed model was validated by simulating by several volume change tests carried out on different bentonites. The results confirmed that the proposed model shows more advantages in describing the THM volumetric behavior.

Gas migration at the granite–bentonite interface under semirigid boundary conditions in the context of high‐level radioactive waste disposal

AbstractThe corrosion of waste canisters in the deep geological disposal facilities (GDFs) for high‐level radioactive waste (HLRW) can generate gas, which escapes from the engineered barrier system through the interfaces between the bentonite buffer blocks and the host rock and those between the bentonite blocks. In this study, a series of water infiltration and gas breakthrough experiments were conducted on granite and on granite–bentonite specimens with smooth and grooved interfaces. On this basis, this study presents new insights and a quantitative assessment of the impact of the interface between clay and host rock on gas transport. As the results show, the water permeability values from water infiltration tests on granite and granite–bentonite samples (10−19–10−20 m2) are found to be slightly higher than that of bentonite. The gas permeability of the mock‐up samples with smooth interfaces is one order of magnitude larger than that of the mock‐up with grooved interfaces. The gas results of breakthrough pressures for the granite and the granite–bentonite mock‐up samples are significantly lower than that of bentonite. The results highlight the potential existence of preferential gas migration channels between the rock and bentonite buffer that require further considerations in safety assessment.

Silica-Reinforced AMP-Calcium Alginate Beads for Efficient and Selective Removal of Cesium from a Strong Acidic Medium.

Due to the significant selectivity for Cs+, ammonium molybdophosphate (AMP) possesses potential to uptake radiocesium from high-level liquid waste (HLLW), whereas its micro-crystalline structure and fine powder morphology limit its industrial application. Although the granulation method with alginate is prospective for the preparation of an AMP exchanger, the mechanical strength of obtained beads may be insufficient for application. In this context, we prepared silica-reinforced AMP-calcium alginate (ACS) beads and evaluated their performance for Cs+ removal from strong acidic solutions. It was found that the addition of silica in the fabrication significantly improved the mechanical strength of the beads in comparison to those without silica. Notably, the beads with an AMP/silica mass ratio of 1.0 exhibited an exceptional mechanical strength, surpassing that of ACS beads composed of other components. The batch experiment results indicated that the Cs+ adsorption follows a non-linear pseudo-second-order rate equation. The distribution coefficient of Cs+ was high even in extreme acidic conditions (∼1.6 × 102 mL/g in 8.0 mol/L HNO3 solution). The Cs+ adsorption can be well fitted with the Langmuir model, and the estimated maximum exchange capacity in 3.0 mol/L HNO3 could reach 23.9 mg/g. More importantly, ACS beads showed excellent selectivity toward Cs+ uptake over eight co-existing metal ions in simulated HLLW, with separation factor values all above 145. The column experiment exhibited that the beads can serve as the stationary phase in columns to effectively remove Cs+. The findings of this study are significant as they provide insights into the development of efficient materials for radiocesium removal from high-level liquid waste. The results demonstrate the potential of silica-reinforced ACS beads for Cs+ adsorption, with promising applications in industrial settings.

Thermal effects on the strain rate-dependent behavior of highly compacted GMZ01 bentonite

Investigation of thermal effects on the strain rate-dependent properties of compacted bentonite is crucial for the long-term safety assessment of deep geological repository for disposal of high-level radioactive waste. In the present work, cylindrical GMZ01 bentonite specimens were compacted with suction-controlled by the vapor equilibrium technique. Then, a series of temperature- and suction-controlled stepwise constant rate strain (CRS) tests was performed and the rate-dependent compressibility behavior of the highly compacted GMZ01 bentonite was investigated. The plastic compressibility parameter λ, the elastic compressibility parameter κ, the yield stress p0, as well as the viscous parameter α were determined. Results indicate that λ, κ and α decrease and p0 increases as suction increases. Upon heating, parameters λ, α and p0 decrease. It is also found that p0 increases linearly with increasing CRS in a double-logarithm coordinate. Based on the experimental results, a viscosity parameter α(s, T) was fitted to capture the effects of suction s and temperature T on the relationship between yield stress and strain rate. Then, an elastic-thermo-viscoplastic model for unsaturated soils was developed to describe the thermal effects on the rate-dependent behavior of highly compacted GMZ01 bentonite. Validation showed that the calculated results agreed well to the measured ones.

Differences and implications of strontium distribution coefficient on various granite compositional materials.

The distribution coefficient (Kd) of radionuclides is a crucial parameter in assessing the safety of high-level radioactive waste (HLW) geological repository. It is determined in the laboratory through batch and column experiments. However, differences in obtained Kd values from distinct experiments have not been thoroughly assessed and compared. This study evaluated strontium (Sr) sorption on different granite materials using static batch and dynamic experiments (column and core-flooding experiments). The results from batch sorption experiments showed higher Sr sorption on granite under acidic and strongly alkaline conditions, low solid-liquid ratios, and low ionic strength. In column experiments, a two-site sorption model was used to simulate Sr transport in crushed granite and mixed pure minerals. The sorption of Sr on crushed granite exhibited a higher affinity than that of mixed pure minerals. The dual-porosity transport model was employed to investigate Sr transport behavior in fractured granite in the core-flooding experiment. Kd obtained from batch sorption experiments are four to twenty times higher than those from column experiments, and two to three orders of magnitude higher than that from a core-flooding experiment. The results of this study provide valuable insights into safety assessment for the HLW geological repository.

Kinetics study on the temperature-dependent reduction of aqueous U(VI) by natural pyrite

Previous studies have indicated that acid-washed pyrite is inert toward aqueous U(VI) under most pH conditions at ambient temperature, even though insoluble UO2 is the thermodynamically predicted product for the reduction of U(VI) by pyrite. Considering the exothermic nature of nuclear waste, the interaction between uranyl nitrate/acetate and natural pyrite was studied at temperatures ranging from 25 °C to 85 °C and at pH values between ∼4.0 and ∼9.5, to simulate the scenarios encountered in a high-level radioactive waste repository. The results revealed that the reactivity of pyrite toward aqueous U(VI) significantly increased with rising temperature, and the reaction could be described by a pseudo-first-order kinetic equation. Activation energies were calculated to be 79.4 ± 10.6 and 45.7 ± 3.8 kJ⋅mol−1 for the reduction of uranyl nitrate, and 78.2 ± 5.2 and 42.2 ± 5.8 kJ⋅mol−1 for the reduction of uranyl acetate, at pH values of ∼4.5 and ∼5.0, respectively. These values indicate that the reaction is controlled by the surface chemical processes. In addition, the complete reduction of aqueous U(VI) to UO2 product was first observed for the reactions at pH ∼4.0 to ∼5.5 when the temperature was ≥75 °C. Conversely, non-stoichiometric UO2+x(s) (0 < x ≤ 0.67) was found at lower temperatures within the same pH range, and also at 85 °C with a reaction pH of ∼9.5. Moreover, the ratio of reduced U(IV/V) on pyrite surfaces increased gradually over time, indicating that reaction time played a significant role in the reduction products. The findings are essential not only for the safety assessment of the high-level radioactive waste repository, but also for understanding the metallogenic mechanism of U(IV)-bearing ore deposits under the relevant anaerobic and hydrothermal conditions.

Analysis of core quality criteria and requirements for high-level waste disposal R&D based on the US NQA-1 and Korea KEPIC-QAP

This study presents the key quality criteria and requirements as well as an implementation plan for R&D related to the disposal of high-level waste. From the results of the analysis based on NQA-1 and KEPIC-QAP, it was found that test control, software control, design control and independent review of the report were derived as QA key criteria. In addition, the R&D organization should gradually expand the quality assurance criteria to increase R&D efficiency for licensing the construction of the spent nuclear fuel repository. In order to improve research quality, it is necessary to customize the quality requirements for the organization. That is, from the 18 quality criteria of NQA-1, such factors as the quality requirements of organization and quality assurance program, design and software control, document control, purchasing management, test control, quality assurance records, and audit can be selected as R&D quality assurance criteria. In other words, high-level waste disposal R&D should be performed considering not only nuclear quality standards but also research resources. However, as of yet, legal R&D quality assurance requirements in Korea are not specific and only general standards are presented.

Swelling characteristics of montmorillonite mineral particles in Gaomiaozi bentonite

Highly compacted bentonite is an ideal back-filling (buffer) material for the deep geological disposal of high-level radioactive waste. Montmorillonite is the main constituent mineral of it. The nanoscopic swelling characteristic of montmorillonite mineral particles greatly influences the macroscopic buffering performance of highly compacted bentonite blocks. Firstly, the montmorillonite suspension liquid was exfoliated by freeze-thaw cycles, the appropriate mineral particles were selected by atomic force microscope (AFM) for making colloidal probes. Then, the swelling pressure between mineral particles was measured by AFM, investigating the effects of interlayer distance, layer number, temperature, and interlayer cation type on the nanoscale swelling pressure. The experimental results showed that the swelling pressure of montmorillonite mineral particles peaks at about 0.7–0.8 nm interlayer distance. The swelling pressure of mineral particles with more layers is smaller. The higher the ambient temperature, the greater the swelling pressure. The swelling pressure of Ca-montmorillonite is significantly larger than that of Na-montmorillonite. The Laird model has a relatively good applicability with the interlayer distance of 0.6–1.2 nm; the DLVO theory has a good applicability with the interlayer distance greater than 1.5 nm. Finally, a predictive model was developed for the swelling pressure of nanoscopic mineral particles considering the effect of layer number. The research results provide data support and theoretical support for swelling pressure generation and cross-scale transmission mechanism of highly compacted bentonite hydration.

Chemical stability and leaching mechanism of YIG and HEG at different pH conditions

AbstractDesigning waste forms using the cocktail effect of high‐entropy ceramics can improve the efficiency of ceramic waste forms. In this work, traditional garnet (Y1.2Nd1.8Fe5O12, YIG) and high‐entropy (HE) garnet (Y0.6Gd0.6Sm0.6Eu0.6Dy0.6Fe5O12, HEG) are synthesized through ignition and plasma sintering to form dense ceramic waste forms. The chemical stability of YIG and HEG was studied by using static leaching tests at pH = 3, 5, 7, 9, 11. The results indicate that acidic environments can increase the leaching rate of ceramics. Leaching behavior leads to a decrease in lattice volume, but the elements remain uniformly distributed. The leaching mechanism indicates that YIG and HEG are controlled by dissolution and surface effect in the early stage (1–7 days) of leaching. The leaching mechanism in the later stage (14–42 days) is characterized by diffusion control. The diffusion coefficient of long‐term leaching indicates that HEG is more suitable as the immobilization substrate for high‐level radioactive waste (HLW) than YIG.

Feasibility of 129I groundwater dating calibrated by both 81Kr and 4He for the assessment of deep geological repositories in Japan

Iodine-129, which is a promising tracer for dating old groundwater, has been used as a tracer for deep upwelling groundwater. The nuclide is expected to be one of the key factors for site selection for high-level radioactive waste disposal, which is a global societal issue. The pre-anthropogenic 129I/127I ratio for marine iodine is (1.50 ± 0.15) × 10−12, which could be considered the initial value for 129I dating. This study identifies the challenges in groundwater age dating using 129I/127I. We measured the ratios of 129I/127I and 81Kr/Kr and concentration of 4He in groundwater from boreholes on the northern coast of Japan. The 129I dating results were not coincident with the other groundwater dating results. The iodine in the groundwater was inferred to be released in situ from marine organisms in sediments of various ages. We estimated that the primordial iodine ratio originating from seawater was ~ 1 × 10–13 (8 × 10–14 ~ 2 × 10–13). The groundwater age deduced from the 129I/127I ratio using this value agrees with other groundwater dating results.

Effects of temperature on fracture and damage characteristics of deep granite

Investigating the fracture- and failure-related behaviors of rock that is subjected to temperature treatment is important for handling warm rock reservoirs during deep mining of hot dry rock and processing high-level radioactive waste. In this study, we use the semi-circular bending test in combination with acoustic emission (AE) monitoring technology to examine the characteristics of fracture and damage in granite treated at different temperatures and under different fracture modes at a depth of 750 m in the Daliuhang Gold Mine in China. The results showed that the peak load and fracture toughness of granite decreased to varying extents when it was treated at increasingly higher temperatures. The high temperature substantially reduced the bonding capacity of the particles of rock, and led to the formation of a large number of microcracks that dislodged the particles of rock along the edges of the samples. The changes in the AE counts during the different loading phases can be categorized into stabilization, increase, sudden increase, and decay stages. The damage-related variable based on the cumulative AE count revealed that samples treated at and below a temperature of 300 °C were mainly damaged in the late period of loading and exhibited brittle failure. Damage began to accumulate as early as in the middle period of loading in samples treated at temperatures greater than or equal to 600 °C. The results of this study provide a useful reference for mining deeply buried granite under different temperature gradients and fracture modes.

Vitrification of Cerium(Ⅲ) fluoride in iron phosphoborate glasses: A study on redox, structure, and stability

Cerium fluoride was vitrified to simulate actinide fluoride waste from high-level waste salt produced through dry reprocessing in a Molten Salt Reactor. Iron phosphoborate and phosphoborate glass were used to immobilize CeF3 and investigate the relationship between redox, structure and chemical stability of these glass wasteforms. The findings indicate that CePO4 crystals precipitated when the molar content of CeF3 exceeded 28.9 wt% in iron phosphoborate glass, as observed by X-ray Diffraction (XRD). X-ray Photoelectron Spectroscopy (XPS) was used to investigate the changes in the valence of Ce and Fe in both types of glass. Results showed that adding of CeF3 increased the fraction of Fe2+. In iron phosphoborate glass, Ce was mainly trivalent, whereas in phosphoborate glass, most Ce was oxidized to Ce4+. Raman analysis indicates that cerium-containing iron phosphoborate glass exhibits a pyrophosphate structure, while cerium-containing phosphoborate glass displays a metaphosphate structure. Results from Mössbauer Spectroscopy and X-ray Absorption Fine Structure (XAFS) indicate that iron predominately exists as Fe–O tetrahedral clusters within the glass, showcasing excellent stability. This is evident from the strong resistance to elemental leaching in the leaching test of the iron phosphoborate glass with cerium content.

The structural, mechanical and chemical stability properties of HEG and YIG in response to α-irradiation

Ceramic materials with excellent radiation resistance stability play an indispensable role in the treatment of nuclear waste. This article uses spark plasma sintering (SPS) to synthesize high-entropy (HE) Y0.6Gd0.6Sm0.6Eu0.6Dy0.6Fe5O12 (HEG) and traditional garnet Y1.2Nd1.8Fe5O12 (YIG). Studied the effects of 2 MeV He2+ irradiation (1 × 1014 ions/cm2 - 1 × 1017 ions/cm2) on the crystal structure, mechanical properties, and chemical stability of ceramics. Research shows that the surface of YIG becomes completely amorphous under ∼30dpa irradiation. Under irradiation at∼30dpa, HEG underwent a small amount of amorphization (with an amorphization rate of 38 %), maintaining the structure of garnet, and the lattice expansion rate of HEG caused by irradiation was lower than that of YIG. After irradiation, the mechanical properties of HEG were improved, while YIG's hardness decreased due to its amorphous state. Radiation has almost no effect on the chemical stability of HEG, and its long-term release mechanism is dominated by diffusion. This study identified HEG as an ideal candidate substrate for immobilizing high-radioactive waste (HLW) the perspective of radiation resistance.

Density spatial distribution and anisotropy of full-scale bentonite-sand blocks

Constructing buffer barriers using high-quality bentonite-based blocks is crucial for the secure disposal of high-level radioactive waste (HLW). However, the density distribution and anisotropy of full-size bentonite-sand blocks are not yet fully understood, which may affect their quality and performance as buffer barriers. This study aimed to address this critical issue by conducting computed tomography (CT) scanning tests on full-size bentonite-sand blocks in the radial, angular, and vertical directions to obtain CT numbers and density. The analysis of the density distribution and anisotropy in these three directions followed. Subsequently, thermal conductivity and hydraulic conductivity tests were conducted in these directions to explore their relationship with density anisotropy. The findings revealed that the block density gradually decreased with increasing radius in the radial direction and first decreased slightly and then increased slightly in the angular and vertical directions. However, the density of the bentonite-sand mixture inside the blocks ranged from 1.96 g/cm³ to 2.09 g/cm³, with minimal differences. Additionally, there was no significant difference in density between the radial and angular directions, with an average density of 2.06 g/cm³. The density in the vertical direction was slightly lower, with an average density of 2.04 g/cm³. Scanning electron microscope images validated the presence of anisotropy in block density caused by horizontally oriented bentonite aggregates resulting from axial compression. However, the presence of quartz sand and incompressible impurities in the blocks created preferential flow paths, mitigating the impact of block density anisotropy on thermal and hydraulic conductivity. Consequently, thermal conductivity, averaging 1.60 W/m·K, and hydraulic conductivity, averaging 3.12 × 10−10 cm/s, demonstrated isotropy in all three directions. These block parameters meet the design requirements for an effective buffer barrier. This study systematically reveals the density distribution of full-size bentonite-sand blocks and its impact on the performance, providing important references for the safe design of HLW repositories.

Long-term tracking of the microbiology of uranium-amended water-saturated bentonite microcosms: A mechanistic characterization of U speciation

Deep geological repositories (DGRs) stand out as one of the optimal options for managing high-level radioactive waste (HLW) such as uranium (U) in the near future. Here, we provide novel insights into microbial behavior in the DGR bentonite barrier, addressing potential worst-case scenarios such as waste leakage (e.g., U) and groundwater infiltration of electron rich donors in the bentonite. After a three-year anaerobic incubation, Illumina sequencing results revealed a bacterial diversity dominated by anaerobic and spore-forming microorganisms mainly from the phylum Firmicutes. Highly U tolerant and viable bacterial isolates from the genera Peribacillus, Bacillus, and some SRB such as Desulfovibrio and Desulfosporosinus, were enriched from U-amended bentonite. The results obtained by XPS and XRD showed that U was present as U(VI) and as U(IV) species. Regarding U(VI), we have identified biogenic U(VI) phosphates, U(UO2)·(PO4)2, located in the inner part of the bacterial cell membranes in addition to U(VI)-adsorbed to clays such as montmorillonite. Biogenic U(IV) species as uraninite may be produced as result of bacterial enzymatic U(VI) reduction. These findings suggest that under electron donor-rich water-saturation conditions, bentonite microbial community can control U speciation, immobilizing it, and thus enhancing future DGR safety if container rupture and waste leakage occurs.

Identification and time evolution of thionyl chloride (SOCl2) radiolysis products

Abstract Innovative solutions are needed to reduce the amount of high-level waste generated by used nuclear fuel recycling strategies to support the widespread adoption of sustainable nuclear fission energy technologies. To this end, a new sulfur chloride-based process has been developed to recycle zirconium alloy-based materials, which make up a significant fraction of high-level radioactive waste. To support the continued development of this process, we present new data on the potential reaction pathways over time of the products arising from the gamma and electron beam radiolysis of neat thionyl chloride (SOCl2). Interrogation of the gamma irradiated liquid by Raman spectroscopy provided more conclusive identification of the SOCl2 degradation products, specifically sulfur dichloride (SCl2), molecular chlorine (Cl2), sulfur dioxide (SO2), and sulfuryl chloride (SO2Cl2). In comparison, the high dose rate (∼107 Gy s−1) electron beam irradiations formed significantly more degradation products. For both cobalt-60 gamma and electron beam irradiations, the observed degradation products were found to evolve as a function of time post-irradiation via the same reaction pathways, with indication of a solvent regeneration mechanism. These findings are fortuitous for process development, as such a mechanism would be beneficial for process longevity and cost effectiveness.

Exploratory neutronic evaluation on the enhancement of tritium breeding and energy multiplication capability using (Th,U)O2 in a solid-state tritium breeding blanket of a fusion demonstration reactor

The achievable tritium breeding ratio (TBR) in Korean fusion demonstration reactor (K-DEMO) that utilizing the deuterium-tritium nuclear fusion reaction for energy production is the critical design parameter for the self-sufficiency of tritium. However, the calculation uncertainty of the achievable TBR makes it difficult to fully understand the tritium fuel cycle in the K-DEMO. Moreover, if the achievable TBR is less than the required TBR, the calculation uncertainty of the achievable TBR makes the design limitation on the self-sufficient tritium supply of the K-DEMO. Therefore, making the achievable TBR larger than the required TBR is the most important goal of the K-DEMO design for the self-sufficient tritium supply. To create an achievable TBR with a margin greater than the required TBR, the modified neutron-multiplying material that is different from the typical neutron-multiplying material (9Be) is tested in this study. The blending of the fissionable (232Th) and fissile (233U) material with 9Be is attempted to enhance the neutron multiplication reaction resulting in a gradual increase of the achievable TBR according to the increase of the atomic faction of fissile material. With the addition of fissile material, it is found that energy multiplication accompanied by fission reaction is dramatic. The possibility of the self-breeding of 233U from 232Th is also justified. However, the neutronic damages resulting in the frequent replacement of the components in the tritium breeding blanket and generation of the high-level waste also occurred as the negative impact accompanied by the blending of 232Th and 233U with 9Be.

Nuclear waste glass alteration under the influence of iron, claystone, and cementitious grout: An integral study

To prepare for the future deep geological disposal of high-level waste, the alteration of two inactive nuclear glasses was studied in the presence of claystone, iron, and cementitious grout at 50 °C for a period of 4 years. The materials were immersed together in a synthetic water representative of claystone solution composition. Blank experiments were also carried out to study the independent behaviors of iron and glass in the synthetic water. The evolution of the solutions’ chemistry was monitored. At the end of the experiments, the glass and iron, as well as the neoformed phases, were characterized. Results showed that Si, Ca, Mg, and Fe are key elements. A competition exists between the retention/integration of species into glass gels and clay-like neoformations. Magnesium tends to be leached from glasses and form Mg-silicates. It would only integrate into the gel if it was widely available in the solution. Furthermore, iron could form Fe-silicates. Whether for Mg or Fe, the formation of silicates was detrimental to the glass since it involved the creation of a silicon sink and thus the conservation of a high alteration rate. Concerning calcium, in these experiments it appeared that this species tends to be integrated/retained within the gels. In the pH conditions existing here, it seems that there is no competition with a neoformation. The specific impact of the cementitious grout was also studied, and the results showed that the presence of this material in the system had a beneficial effect on glass alteration due to a significant Si supply in the solution, enabling a reduction in the glass alteration rate.

In situ heating high-resolution TEM observation of structural recovery in metamict titanite

Besides radiation resistance, thermal recovery of radiation damage provides another critical criterion to evaluate the host phase for immobilizing high-level waste (HLW) produced by nuclear plants. Despite the increasing investigations, a comprehensive atomic-scale understanding remains elusive on thermal annealing of radiation damage in ceramics. Here, using in situ heating high-resolution transmission electron microscopy, we investigate the thermal annealing of radiation damage in natural metamict titanite (CaTiSiO5) caused by a high alpha decay dose from the incorporated U and Th during the geologic time. The recovery of the fully amorphous titanite precursor starts at 456 ℃ from the appearances of monoclinic nanoparticles 2–5 nanometers wide, and more nanoparticles form as the temperature gradually increases. At 1000 ℃, the amorphous phase recovers through epitaxial growth with titanite nanoparticles as templates. This study provides information for a unit-cell-scale understanding of the thermal stability of radiation damage in titanite-based ceramics, having significance for immobilizing HLW.

Precise separation and efficient recovery of Pd(II) from high-level liquid waste by XAD-based adsorbents

Precise separation and efficient recovery of Pd(II) from high-level liquid waste by XAD-based adsorbents