High-level Nuclear Waste Research Articles

Extraction of Ln(III) and An(III) by N,N-di(2-ethylhexyl)- thio-diglycolamic Acid

ABSTRACT The extraction of Nd(III), Eu(III), Am(III), and Cm(III), trivalent lanthanides and actinides in high level nuclear waste, by the extractant N,N-di(2-ethylhexyl)-thio-diglycolamic acid (HDEHSDGA, HA) is investigated via thermochemical and spectroscopic methods. In this study, HDEHSDGA behaves as a conventional mono-basic cation-exchanging ligand extracting Ln(III) and An(III) from nitrate solutions of pH region. The information obtained from the extraction experiments and the spectroscopic analyses indicate that the primary extracted complex species may be MA3 · 2H2O for these four cations and likely for the other trivalent f-elements. In the extracted complexes, the anionic monomer extractant A− coordinates to the metal ions presumably forming bidentate coordination via the carbonyl oxygen atom and one of the oxygen atoms of the carboxylate; it appears that the thio-ether does not directly participate in the complexation reaction. The results ultimately demonstrate that HDEHSDGA does not exhibit significant selectivity for An(III) over Ln(III), in contrast to results reported in a previous study conducted with unpurified extractant.

Management of Radioactive Waste from HTGR Reactors including Spent TRISO Fuel—State of the Art

In light of the increasing demand for energy sources in the world and the need to meet climate goals set by countries, there is growing global interest in high temperature gas cooled reactors (HTGRs), especially as they are known to be inherently safe nuclear reactors. The safety of HTGRs results, among other, from the nature of the nuclear fuel used in them in the form of coated TRISO particles (tri-structural-isotropic) and the reduction of the total amount of radioactive waste generated. This paper reviews numerous methods used to ensure the sustainable, feasible management and long-term storage of HTGR nuclear waste for the protection of the environment and society. The types of waste generated in the HTGR cycle are presented as well as the methods of their characterization, which are important for long-time storage and final disposal. Two leading nuclear fuel cycle strategies, the once-through cycle (direct disposal or open cycle) and the twice-through cycle (recycling or partially closed cycle), are discussed also in relation to TRISO spent fuel. A short review of the possibilities of treatment of TRISO spent nuclear fuel from HTGR reactors is made.

The impact of 40 years of radiation on the integrity of copper

Cu piping previously used to carry process air was harvested from the now shutdown National Research Universal reactor at the Canadian Nuclear Laboratories site in Chalk River, Ontario. Metallurgical and microscopy examinations found no quantifiable differences between irradiated and non-irradiated sections of the pipe. This provides evidence for the safety case for the use of the Cu-coated used fuel container proposed by the Nuclear Waste Management Organization for the permanent disposal of high-level nuclear waste.

Development of triaxial compressive apparatus for neutron experiments with rocks.

Underground engineering for processes such as geological disposal of high-level nuclear waste, CO2 capture and storage, and mining and drilling for resources requires an understanding of the mechanical behavior of rocks at subsurface stress states, i.e., triaxial compressive stress. Strain measurement using neutron diffraction can be applied to rocks to analyze strain accumulation mechanisms at the microscopic scale. This study reports the development of triaxial compressive apparatus for strain measurement using neutron diffraction. The apparatus can analyze rock specimens (diameter, 25mm; length, 50mm) and apply a maximum confining pressure of 50 MPa. Materials for the components of the apparatus were investigated theoretically based on neutron beam transmission and experimentally using neutron diffraction experiments. The feasibility of the apparatus was verified by measuring strain at hydrostatic pressure under the application of confining pressure and triaxial compression. The theoretical and experimental results show that the apparatus could obtain sufficient neutron statistics from a rock specimen. It was confirmed experimentally that the measured strain values are correlated with the applied confining pressure and stress. The lattice strains of quartz minerals measured by neutron diffraction showed linear deformation behavior, indicating that elastic strain accumulated in the minerals. This apparatus will enable the finding of new insights into the deformation mechanisms of rocks.

Effect of Al and Fe sources on conversion of high-level nuclear waste feed to glass

• Glass batch (melter feed) components affect melting rate through enthalpy and foaming. • Alumina sources affect silica dissolution, glass melt formation, and foam generation. • Gibbsite particle size controls glass production rate through affecting melt foaming. • Chemically bound water in Al and Fe sources affects the glass production rate. Pilot-scale melter testing has shown that Al and Fe sources in high-level nuclear waste feeds influence the glass production rate. To examine melting behaviors of these feeds, we employed X-ray diffraction, differential scanning calorimetry, thermo-gravimetric analysis, particle size analysis, and the feed expansion experiments. Both the chemical form and the particle size of the Al and Fe sources affect the rate of melting through their effects on the conversion enthalpy and the primary foam formation that control the energy demand for melting and the heat accepted by the cold cap. The particle size of gibbsite controls the rate of alumina incorporation in the initial glass-forming melt that in turn, through its effect on viscosity, affects the rate of dissolution of silica particles, thus governing the glass melt fraction, open pore closure, and primary foam formation. Since the temperature of primary foam collapse limits the heat flow to the cold cap, the gibbsite (and generally Al source) particle size ultimately influences the glass production rate that increases as the particle size increases. Variation in the Fe source affects the glass production rate mainly through their content of chemically bound water.

Soil–water characteristic curves and hydraulic conductivity of Gaomiaozi bentonite pellet-contained materials

The bentonite pellet-contained material (PCM) is a desired joint sealer for high-level radioactive waste repositories. During the operation of the repository, however, the PCM is unsaturated for most of the period. The water retention capacity and permeability of PCM can thus directly affect the bentonite buffer barrier seepage, nuclide migration, and joint healing. Moreover, the performance of PCM can be strongly influenced by the particle size of bentonite pellets and dry density. In this work, the pressure plate method and vapor equilibrium technique were both utilized to test the soil–water characteristic curves (SWCCs) of the PCMs with varying particle sizes and dry densities. The unsaturated hydraulic conductivity of the PCMs was then predicted by combining the SWCC model and saturated hydraulic conductivity. The results showed that in the low suction range (20–1150 kPa), the dry density and particle size are anti-correlated with the water content. In the high suction range (4200–309,000 kPa), the SWCC is almost insensitive to the variation of dry density and particle size. The Gardner model is suitable for describing the SWCC of PCM. In addition, the hydraulic conductivity of the PCM increases with the rise in particle size, while it decreases as the dry density grows. Finally, the influence mechanism of the SWCC and hydraulic conductivity was further discussed based on the scanning electron microscopy images and pore size distribution curves.

The role of trust and risk perception in current German nuclear waste management.

One of the lessons learned in various countries that have to deal with spent nuclear fuel is that finding a proper place and siting a repository for high-level nuclear waste (HLW) cannot be achieved without public consent. After decades of obstruction, Germany recently launched a new, participatory, site-selection process for the disposal of HLW in deep geological formations. Nonetheless, significant opposition is assumed. Therefore, citizens' trust in the procedure and the agents involved may be paramount. We conducted an online survey (N ≈ 5000) in March/April 2020 to test a theoretical model on trust, perceived risks and benefits, and acceptance. We differentiated acceptance as a dependent variable according to distinct phases: the procedure, a possible decision on a disposal location, and the repository facility itself. The results show that trust is mainly important for explaining acceptance of the ongoing procedure and less so for the acceptance of the decision or the repository facility itself. Moreover, our investigation of the sample using a cluster analysis reveals characteristic patterns of trust, risk perception, and acceptance by three clusters: a cluster focusing on risk perception, an ambivalent cluster, and an indifferent cluster. Trust is lowest in the risk-focused cluster and highest in the ambivalent cluster.

Analyzing of the hydration crack evolution in compacted GMZ bentonite with consideration of technological void ratio

Technological voids widely exist around the bentonite blocks in the deep geological repository. After groundwater seeps into the voids, bentonite nearby expands to fill up the voids and completes the self-sealing gradually. The self-sealing ability is of highly importance, which prevents the technological voids from playing a preferential channel for groundwater migration and thus threatening the safety of the canister accommodating the high-level radioactive nuclear waste (HLW). So far, much of the research focus has been on the hydro-mechanical properties of bentonite after self-sealing, while the process itself is neglected. But actually, a cracking phenomenon will pop up in the interim of sealing process. In this paper, the dynamic evolution of self-sealing on compacted bentonite with different technological voids and sample thicknesses was observed by using a transparent observation apparatus. Digital image processing method combined with fractal theory was utilized to process, analyze and quantify the images captured. The results show that the self-sealing process contains the lifetime of the hydration crack, including its appearance, development and extinction. The technological void ratio evidently affects the time require for closing stage compared with the cracking stage. But in the range of 10–25%, there is no clear correlation between the technological void ratio and the geometry parameters or complexity of crack network. The fractal dimension introduced to quantify the complexity corresponds well with the visible image result, showing the feasibility and credibility of box-counting method. The cracking and closing mechanisms are also investigated.

Co-transport of bentonite colloid and U(VI) in particulate granite column: role of colloid concentration, ionic strength, pH and flow rate

Abstract China is considering Beishan granitic formation (Gansu Province, China) as the site for high-level radioactive waste (HLW) repositories. Thus, it is crucial to understand the transport behavior of radionuclide in Beishan granitic media under disposal conditions. In this context, the co-transport of U(VI) (as the representative of radionuclides) and bentonite colloid (BC, from erosion of compacted bentonite) in particulate Beishan granite was studied as a function of important in-situ factors, such as BC concentration, ionic strength, pH and flow rate. We found that the increase of BC concentration (BC = 240–480 mg/L) did not affect the transport of individual BC, whereas it significantly facilitated the transport of U(VI). The increase of ionic strength (I = 0.001–0.01 M NaCl) or decrease of pH (pH = 7.50–5.40) obviously inhibited the BC transport, where these inhibiting effects were relatively slight for the transport of U(VI). The increase of flow rate significantly facilitated both the transport of BC and U(VI). Finally, a two-site kinetic attachment/detachment model was applied to describe the breakthrough curves of individual and co-transport of BC. The experimental and modeling results of this study have a significant implication on the safety assessment of HLW repositories built in granitic formation.

Coupled THM analysis of a full-scale test for high-level nuclear waste and spent fuel disposal under actual repository conditions during 18 years of operation

This study focuses on the thermo-hydro-mechanical (THM) analysis of the behaviour of a full-scale engineered barriers experiment designed to study the long-term response of a disposal concept for high-level radioactive waste (HLW) and spent fuel (SF). Crucially, the test involved two dismantling operations, a partial one after 5 years of heating and a final one, after 18·3 years of operation, allowing the direct observation of the state of the barrier at two different times. The coupled THM modelling discussed in this paper mainly focuses on the last 13 years of test operation and also on the final dismantling of the barrier system. The numerical model was unchanged from the early stages of the test, so the results presented can be considered as blind predictions. The numerical model predicted reasonably well the state of the barrier and its evolution, especially in terms of the distribution of the bentonite dry density, confirming the predictive capability of the model to capture the transient THM processes. The combination of experimental and numerical investigations, involving one of the most long-lasting full-scale in situ experiments monitoring the behaviour of the barrier components of an HLW/SF repository, has led to a better understanding of the THM behaviour of this complex system.

The process to find a process for governance: Nuclear waste management and consent-based siting in the United States

In 2013, the United States (U.S.) Department of Energy (DOE) attempted to initiate a consent-based siting (CBS) approach to better engage diverse publics and thereby begin to remedy a legacy of technocratic decision-making and inequitable public engagement processes plaguing historical high-level nuclear waste siting efforts. DOE's remediation work included a contract with the Expert and Citizen Assessment of Science and Technology (ECAST) network to employ participatory technology assessment (pTA) to help co-create the CBS process with lay publics. In late 2016, the DOE terminated the process to develop a CBS process. To date, little in the public record explores DOE's novel foray into CBS. As researchers on DOE's aborted ECAST efforts, we situate the novel pTA process to create a CBS process amidst the technocratic political-historical context of commercial nuclear waste siting in the U.S. Lessons from ECAST's effort highlight persistent institutional barriers inhibiting U.S. capacity to more equitably approach the challenge of siting nuclear waste facilities. We identify the undermining consequences of DOE's focus on expediency; imposed limitations on the scope of CBS; bureaucratic obstacles to public input; a lack of continuity in values across executive administrations; and absence of top-level commitment to procedural and institutional learning, innovation, and adaptation. Through our case history and critical reflection, we aim to inform future efforts in the U.S. and beyond to overcome failed technocratic histories and instead steward participatory, equitable, and democratic processes to manage high-level nuclear waste.

Conversion kinetics during melting of simulated nuclear waste glass feeds measured by dissolution of silica

We experimentally investigated the batch-to-glass conversion kinetics of two low-activity and two high-level nuclear waste melter feeds that represent widely different behavior during melting; tests comprised thermogravimetric analysis, the feed expansion test, evolved gas analysis, and x-ray diffraction. Our study indicates that the conversion processes, which include gas-evolving reactions, dissolution of crystalline phases, and primary foam expansion and collapse, are interconnected and influenced by the feed's thermal history. The fraction of dissolved silica was adopted as a measure of the extent of conversion inside the cold cap—the floating layer of feed material on the top surface of the melt in an electric melter. The interface between the cold cap and the free-flowing melt below was associated with a constant fraction of dissolved silica. The estimated viscosity of glass-forming melt at this interface lies within a narrow range for the feeds studied.

Survival and activity of an indigenous iron-reducing microbial community from MX80 bentonite in high temperature / low water environments with relevance to a proposed method of nuclear waste disposal

MX80 bentonite clay has been selected as the buffer and backfill in a proposed method for long-term deep geological storage of nuclear waste. Extensive studies have been carried out on the geomechanical properties of the clay; however, it is not clear what effect microbes, specifically iron-reducing bacteria, will have on its ability to function as an affective barrier. Iron-reducing bacteria can reduce structural or external Fe(III) to Fe(II) and have been previously identified in the indigenous microbial community of MX80 bentonite. Experiments to assess bacterial survival at the high temperature and low water conditions likely to exist in the repository were carried out at different temperatures with the addition of steel to represent a nuclear waste canister. The resulting microbial enrichments were analysed, and mineralogical and geomechnical analysis was carried out on the clay. Microbial sequencing revealed that iron-reducing bacteria, and other indigenous species can survive these conditions in MX80 bentonite in either an active or dormant state. Microbial influenced mineralogical changes may lead to a loss of silica from the clay and reduction of Fe(III) to Fe(II). These changes could alter the ability of the clay to act as an effective barrier in nuclear waste disposal. Furthermore, evidence of reduced steel corrosion when microbes were present suggested that microbial activity may lead to either a protective coating on the steel or depletion of oxygen to slow corrosion. The production of such a layer would benefit nuclear waste disposal by inhibiting corrosion of a metal waste canister.

Strengthened erosion resistance of compacted bentonite by layered double hydroxide: A new electrostatic interaction-based approach

For the geological repository of high-level radioactive waste (HLW) built in granitic host rock，the control of buffer material (compacted bentonite) erosion and subsequent loss caused by groundwater in granite fissures is an unresolved problem of major concern. We propose here new insight into enhancing the erosion resistance of compacted bentonite by means of its electrostatic interaction with oppositely-charged layered double hydroxide (LDH). The interaction between bentonite and LDH was studied by dropwise addition of colloidal LDH into colloidal bentonite suspension, during which the variation in electrical conductivity, zeta potential and particle size proved a strong interaction between these two materials. Interestingly, in addition to their aggregation, intercalated structures of LDH and montmorillonite were found in the composite (BEN@LDH) by a combined characterization of X-ray diffraction (XRD) and high resolution transmission electron microscopy (HR-TEM), and were confirmed by density functional theory (DFT) calculation. Colloid generation of compacted BEN@LDH under ultrasonic conditions is negligible comparing with that of compacted bentonite, indicating a significantly higher erosion resistance. Besides, a small amount of LDH by mechanically mixing with bentonite (mass ratio 1:99) can also effectively improve the erosion resistance of compacted bentonite. Moreover, BEN@LDH displayed stronger retention performance towards U(VI) and Se(IV) than bentonite under near-neutral/weakly alkaline conditions. Our results indicate that LDH is a promising additive in compacted bentonite, and this approach may be extended to common geotechnical structures built with clays and soils.

Aggregated germanium saponite: Removal and retention of polymeric thorium and uranium complexes

The present study concerns the removal of thorium and uranium from aqueous solution, and particularly the effect of clay particle size and morphology on the adsorption of polynuclear actinide complexes by synthetic saponite nanoscale powders. Saponites were synthesized from a reactant mixture containing aluminum nitrate, magnesium nitrate, and either sodium silicate or sodium germanate. The saponite powders were analyzed by TEM, XRD, SAXS, N2-adsorption-desorption isotherms and FT-IR. Ge-saponite contained randomly stacked platelets ≈ 100 nm long and 80 nm wide, while Si-saponite showed roughly spherical agglomerates ≈ 60–80 nm in size. Small-angle X-ray scattering measurements indicated a hierarchical pore network among the plates and their agglomerates. Thorium polymers composed of fractal aggregates of branched filaments were characterized by synchrotron radiation-SAXS. The distribution coefficient (Kd mL/g) values for thorium (20 μg/mL) at pH 4 were: Ge-saponite (105) ≈ Si-saponite (105) > raw saponite (104) ≈ raw saponite (washed) (104). The Kd values for uranium (20 μg/mL) at pH 5.5 were: Ge-saponite (103) ≈ raw saponite (washed) (103) ≈ Si-saponite (103) ≈ raw saponite (103). The difference in uranyl adsorption performance observed between the two nanoparticle forms of saponite was attributed to a larger pore width for Ge-saponite (3.5 nm) compared with Si-saponite (0.8 nm). The results have implications for mineral ore tailings containing thorium(IV) and, by analogy, Pu(IV) in a clay buffer environment within a geological repository for high-level nuclear waste.

Effect of manganese on the speciation of neptunium(V) on manganese doped magnetites

Speciation and fate of radionuclides in environment are strongly dependent on their interaction with various mineral surfaces. Presence of impurity atoms in minerals modifies these interactions. 237Np, a long-lived component of nuclear high-level waste, is highly mobile in environment as neptunyl (NpO2+) cation. Role of impurities in defining its interaction with mineral surfaces present in natural and engineered systems is yet to be completely understood. Manganese is a ubiquitous impurity present in naturally occurring iron corrosion product, magnetite. Here, a series of manganese (Mn)-doped magnetites (FM) was synthesized and the evolution of solid-sorbed speciation of Np(V) was investigated as a function of Mn doping (3–11 wt%) in near neutral pH condition. Combining Mn K-edge and Np L3-edge X-ray absorption fine structure spectra with Np M5-edge high energy-resolution X-ray absorption near-edge structure spectra, we found nanoparticulate NpIVO2 solid as the single speciation of Np sorbed on both doped and undoped magnetites. Without changing the surface speciation of Np, Mn-O sites did show preference for sorbing Np(V) vis-á-vis Fe-O sites. Mn participates through Fe(II)/Mn(III) redox couple in reducing sorbed Np(V). Its participation, however, becomes significant only when the doped-magnetite contains a critical minimum concentration (> 6 wt%) of Mn. This study highlights the significant role of impurity atom wherein it interacts with sorbing species through redox process.

Numerical analysis of safety of a borehole repository for vitrified high-level nuclear waste

We developed a model and carried out transient 3-D computer simulation of radionuclides migration from an underground repository for vitrified high-level radioactive waste (HLW) with long-lived actinides. The repository represents an array of deep boreholes. Canisters with vitrified HLW are disposed one above another in lower parts of the boreholes. The upper parts of the boreholes are sealed. Radionuclides leached from the disposed waste are carried away by thermal convection of groundwater caused by heat generation in HLW. The computer simulation was carried out with use of a refined technique taking into account significant difference between radionuclides concentration in the groundwater in the loaded parts of the boreholes and at the earth surface, dependence of the groundwater properties on pT-parameters, composition of HLW. Values of leaching rate of the waste forms by groundwater depending on time and temperature, as well as role of colloid facilitated transport of radionuclides were determined from previous experiments with Na–Al–P-glass which was used in Russian Federation for HLW vitrification. The results of the calculations allowed us to conclude that distance between containers with HLW and the ground surface of 240 m and more is sufficient for safe disposal of the considered HLW.

Investigation of Re(VII) diffusion in Tamusu clayrock core by through-diffusion method

Abstract Tamusu area is the primary pre-selection site of clayrock disposal repository for high-level radioactive waste (HLW) in China. However, the research on the migration behavior of nuclides in Tamusu clayrock is still in its infancy. For the first time in laboratory, the diffusion behavior of Re(VII) in Tamusu clayrock core was studied by means of through-diffusion method. The effects of pH, ionic strength and humic acid on the diffusion behavior of Re(VII) in clayrock were studied. The effective diffusion coefficient, apparent diffusion coefficient and rock capacity factor value were obtained. All the experimental conditions of Re(VII) diffusion in Tamusu clayrock are compared with other geological samples under the same conditions in literature data. The diffusion mechanism of radionuclide in Tamusu clay is discussed, which can provide experimental data for site selection and safety assessment of high-level radioactive waste repository in China. The experimental results showed an effective application and reference for the countries disposed HLW in mudrocks or clayrocks, such as France, Belgium etc. in Europe. Moreover, this research can provide the original data support for the metallogenic regularity and prospecting prognosis of rare element rhenium in different geological environments.

Thermoplastic cohesive fracturing model of thermally-treated granite

Thermoplastic fracturing is a prominent characteristic of thermally-treated granite for temperature-dependent underground geotechnical engineerings , such as geothermal energy exploration and high-level nuclear waste disposal. More specifically, the tension-open of the cohesive crack in the fracture process zone (FPZ) presents a plastic softening feature dependent on temperature. However, understanding granite thermoplastic fracturing remains challenging. This work proposes a thermoplastic cohesive fracturing model of thermally-treated granite, incorporating temperature-dependent softening rule and completely coupled stress-strain-temperature constitutive relation. Several new thermoplastic fracturing properties in the proposed model are determined by further analyses of published fracturing test data (Miao et al., 2020) done on Beishan granite suffering high-temperature treatments (200–500 °C). The initial yield parameter (cohesive tensile strength) decreases linearly with the rising high-temperature treatments. The critical yield parameter (the critical crack opening displacement (COD)) follows a two-stage linear increasing rule with 300 °C as a boundary, similar to the temperature-dependent FPZ length. The plastic modulus, delineating the cohesive crack's softening rule, increases quadratically with the enhanced high-temperature treatment. The temperature sensitivity modulus ( T ), characterizing the contraction/expansion of yield surface with increasing temperatures, decreases from positive to negative with the increasing COD. Both the fracture energy and the accumulated dissipated energy present remarkable nonlinearities with several inflection points , implying the complex temperature-dependent fracture resistance . This proposed model was validated with three-point-bending fracturing tests of thermally-treated Beishan granite, fitting well (coefficient: 88.8%–99.7%) with the experimental cohesive tensile strength , critical COD, and the accumulated dissipated energy. Besides, by analyzing the softening curves of cohesive cracks in thermally-treated granite, the positive-negative change of T was consistent with the contraction-expansion transition of the yield surface. This work provides a modeling approach for temperature-dependent fracturing of rock-like materials. • The thermoplastic softening of FPZ is characterized via a thermoplastic framework. • This work proposes a thermoplastic cohesive fracturing model for rocks. • New model parameters were proposed to characterize temperature-dependent fracturing. • The correlation coefficient between the model and test data reaches 88.8%–99.7%. • This work provides a modeling approach for temperature-dependent fracturing of rock.

Research Progress of Immobilization of High-Level Nuclear Waste by Using Phosphate Glass-Ceramics

Research Progress of Immobilization of High-Level Nuclear Waste by Using Phosphate Glass-Ceramics