Repository For Disposal Of Nuclear Waste Research Articles

Thermodynamic description of U(IV) solubility and hydrolysis in chloride systems: Pitzer activity model for the system U4+–Na+–Mg2+–Ca2+–H+–Cl––OH––H2O(l)

This study presents updated chemical, thermodynamic, and activity models for the system U4+–Na+–Mg2+–Ca2+–H+–Cl––OH––H2O(l) derived using the Pitzer formalism and a strict ion interaction approach. The models build on comprehensive solubility datasets in dilute to concentrated NaCl, MgCl2, and CaCl2 solutions. The Nuclear Energy Agency-Thermochemical Database (NEA-TDB) selection of solubility and hydrolysis constants in the reference state were taken as anchoring point, and were extended further with the solid nanocrystalline phase UO2∙H2O(ncr) and the ternary complex Ca4[U(OH)8]4+. The former was identified in long-term solubility experiments at ambient conditions, whereas the latter has been selected in analogy to Th(IV), Np(IV), and Pu(IV) considering experimental evidences available for these An(IV) in alkaline, concentrated CaCl2 solutions. These models represent an improved tool for the calculation of U(IV) solubility and aqueous speciation in a variety of geochemical conditions including concentrated brine systems relevant in salt-based repositories for nuclear waste disposal.

Beryllium solubility and hydrolysis in dilute to concentrated CaCl2 solutions: thermodynamic description in cementitious systems

The solubility and hydrolysis of Be(II) was investigated from undersaturation conditions in alkaline, dilute to concentrated CaCl2 solutions (0.05–3.5 M). Experiments were performed with α-Be(OH)2(cr) under Ar atmosphere at T = (22 ± 2)°C. Aqueous Be speciation was further investigated by means of molecular dynamics (MD) calculations. For the most diluted CaCl2 systems (0.05 and 0.25 M), a solubility minimum is observed at pHm ≈ 9.5 {with [Be(II)] ≈ 10−7 M}, consistent with solubility data previously reported in NaCl and KCl solutions. Above this pHm, and at higher CaCl2 concentrations, a steep increase in the solubility with a slope of ∼ +2 is observed, hinting towards the predominance of the moiety [Be(OH)42–] in the aqueous phase. In NaCl and KCl systems, this hydrolysis species prevails only above pHm ∼ 13, thus supporting the formation of ternary complex/es Ca–Be(II)–OH(aq) in CaCl2 solutions. The analysis of solubility data in combination with MD calculations underpin the key role of the complex Ca2[Be(OH)4]2+ in alkaline to hyperalkaline systems containing Ca. In combination with our previous work in NaCl–NaOH and KCl–KOH systems, complete chemical, thermodynamic and (SIT) activity models are derived for the first time for the system Be2+–Ca2+–Na+–K+–H+–Cl––OH––H2O(l). This model provides an accurate and robust tool for the evaluation of Be(II) solubility and speciation in a diversity of geochemical conditions, including source term calculations of beryllium in the context of repositories for nuclear waste disposal with a high cement inventory.

Study on Damage Behavior and Its Energy Distribution of Deep Granite at High-Temperature Conditions

The phenomenon of surrounding rock damage and rupture caused by high temperatures is widespread, and has become a potential threat to the safety of nuclear waste disposal repositories. In order to reveal the energy distribution pattern of fractured granite during the failure process under different confining pressures, triaxial compression tests were carried out on rocks with different initial thermal damage. Firstly, the rock was treated at a high temperature to analyze the change rule of the porosity of the rock after high-temperature treatment, define the equivalent damage coefficient, and analyze the influence of confining pressure and equivalent damage coefficient on the peak stress and peak strain of the rock. The results show that, after high-temperature treatment, the porosity increases with the increase in temperature. The peak stress and corresponding strain of rock samples with similar equivalent damage factors increase with the increase in confining pressure. By comparing the rock samples with the same confining pressure and different initial thermal damage, the larger the confining pressure, the smaller the difference of peak stress of different initial thermal damage specimens. Then, the energy density of rock in a triaxial compression test is quantitatively analyzed by energy theory. The results show that, as long as the confining pressure is the same, the proportion of the dissipated energy of the specimen has a similar evolution trend with the strain. When the confining pressure is the same, the proportion of dissipated energy decreases rapidly with the change of strain due to the increase in equivalent damage factor, but the rate of decline will gradually slow down; however, when the confining pressure increases, the difference caused by the equivalent damage factor will gradually decrease, because the fracture is bound by the confining pressure. Finally, we analyze the maximum dissipated energy during rock deformation and failure. According to the inflection point of maximum dissipated energy, the optimum time for critical support of the key rock mass is determined.

Coupled thermal-hydro model tests of double-layered buffer for nuclear waste repository

In the high-level radioactive waste disposal repository, bentonite is a preferred buffer material. As the buffer material, the hydraulic conductivity must be less than 10−13 m/s and the thermal conductivity should be higher than 0.8 W/(m·K) recommended by International Atomic Energy Agency (IAEA). Due to the decay heat of radionuclide and low thermal conductivity of bentonite, the released heat cannot be quickly transferred to the host rock through bentonite buffer layer and the repository may be in a high temperature state (>100 °C), which compromises both the hydro-mechanical performance of the bentonite and the anti-corrosion behavior of nuclear waste canister, therefore the barrier function of the buffer layer is severely damaged. Thus, a double-layered buffer concept which is based on the dual requirements of thermal conductivity and permeability was proposed for the nuclear waste disposal repository. In this paper, coupled thermal-hydro model tests of double-layered buffer with graphite-bentonite mixture layer and bentonite layer (the layer thickness ratio of graphite-bentonite to bentonite was 2:3) were carried out to investigate the temperature and relative humidity distribution laws and determine whether the physicochemical properties of double-layered buffer change after experiencing the coupled thermal-hydro process. In the model test, a heater simulating the decay heat generated by HLW was placed below the graphite-bentonite layer, and a hydration system simulating groundwater infiltration was installed above the bentonite layer, thus, the heat was transferred from the graphite-bentonite layer to the bentonite layer while water was permeated from the bentonite layer to the graphite-bentonite layer. The temperature and relative humidity distribution were synchronously measured by temperature-humidity sensors. After the model tests, the heater was switched off, and the tested bentonite/graphite-bentonite mixture samples were taken out for subsequent water content, scanning electron microscopy (SEM) and X-ray diffraction (XRD) tests. The main conclusions from the test results are as follows. The distributions of temperature and relative humidity in the buffers were mainly governed by the competition between the thermal effect and the water permeation effect, the results of water content at different distances also indirectly verify this conclusion. And SEM results showed that macropores of the bentonite close to the water inlet side were significantly reduced and the pore distribution tended to be uniform along the water injection direction during wetting. XRD results indicated that the mineral composition of the graphite-bentonite mixtures close to the heater changed while that of the bentonite away from the heater remained unchanged. Obviously, even if the high thermal conductivity graphite was added to the buffer layer near the heat source, the mineral composition of the bentonite at the heat source side cannot be avoided completely unaffected. Therefore, it is necessary to conduct heat from the inner layer to outer layer to reduce the overall temperature in the repository and reduce adverse effects of temperature and water synchronization on outer bentonite layer. The research results in this paper can serve as a guideline for the buffer layer design in the high-level radioactive waste repository.

Experimental research on temperature field construction of nuclear waste barrel based on RDTS with improved spatial resolution

In this paper, a 3D temperature field construction method based on Raman Distributed Temperature Sensor (RDTS) with improved spatial resolution is proposed to monitor the side surface temperature of Nuclear Waste Barrel (NWB). This method includes two aspects: the spatial resolution enhancement of RDTS and the signal dimension conversion. Firstly, the spatial resolution improvement method combining Fully-Connected Neural Network (FCNN) and Total Variation Deconvolution (TVD) is used to improve the spatial resolution of RDTS from 1.6 m to 0.4 m. Then, the one-dimensional temperature signal is converted into 2D temperature field by matrix laying method. Finally, by establishing the coordinate system of the NWB, the 2D temperature field can be converted into a 3D one. The experimental results of the simulation device showed that when the circular thermal area with diameter of 0.4 m, 0.6 m and 0.8 m appeared on the side surface of NWB, the temperature accuracy of the temperature field image constructed by this method can reach 80.91 %, 72.09 % and 85.97 % respectively, and the shape was accurate. The method proposed in this paper can accurately present the surface temperature field of NWB and provide technical support for the safe operation of the nuclear waste disposal repository during its life cycle.

Hydromechanical Performance of Heat-Treated Bentonite

Abstract As a buffer material in nuclear waste disposal repository, bentonite experiences long-term persistent heating because of thermal energy released from the sealed nuclear waste canisters. There is no conclusive evidence on how high temperature affects the self-sealing capacity of bentonites. To investigate this effect, bentonite was heated in a muffle furnace at a constant temperature (105°C) for different time periods. The free swelling ratio, swelling pressure, permeability, and specific surface area are used to characterize the physical and mechanical properties of the heated samples. Results revealed that with the increase in the heating time, the free swelling ratio and swelling pressure decrease significantly. The X-ray diffraction pattern of the bentonite sample heated for 90 days reveals a shift from 6° to 9° for the angle corresponding to the (001) diffraction peak of montmorillonite, and with the increase in the heating time, the spacing distance decreases from 1.5 to 0.94 nm. Meanwhile, siliceous oxide is formed during heating, resulting in a cluster of bentonite particles. Results obtained from the infrared spectroscopy and thermogravimetric analysis revealed that the water adsorbed on the intralayer and interlayer surfaces is completely evaporated when the temperature approaches 200°C, but only part of the ionic-bonded water vaporizes, while the structural water remains unchanged. Furthermore, during evaporation, sodium and magnesium ions disappear. Scanning electron microscopy images revealed that bentonite particles are aggregated, rather than uniformly dispersed, which confirms the aforementioned inference. In addition, results obtained from the particle size distribution tests of heated bentonite revealed the same conclusion about its microstructure.

A Review of Geochemical Modeling for the Performance Assessment of Radioactive Waste Disposal in a Subsurface System

Radionuclides are inorganic substances, and the solubility of inorganic substances is a major factor affecting the disposal of radioactive waste and the release of concentrations of radionuclides. The degree of solubility determines whether a nuclide source migrates to the far field of a radioactive waste disposal site. Therefore, the most effective method for retarding radionuclide migration is to reduce the radionuclide solubility in the aqueous geochemical environment of subsurface systems. In order to assess the performance of disposal facilities, thermodynamic data regarding nuclides in water–rock systems and minerals in geochemical environments are required; the results obtained from the analysis of these data can provide a strong scientific basis for maintaining safety performance to support nuclear waste management. The pH, Eh and time ranges in the environments of disposal sites cannot be controlled, in contrast to those under experimental conditions in laboratories. Using a hypothetical error mechanism for the safety assessment of disposal sites may engender incorrect assessment results. Studies have focused on radionuclide reactions in waste disposal, and have offered evidence suggesting that these reactions are mainly affected by the geochemical environment. However, studies have not examined the thermodynamics of chemical reactions or interactions between water and minerals, such as the surface complexation and adsorption of various nuclide-ion species. Simple coefficient models have usually been applied in order to obtain empirical formulas for deriving Kd to describe nuclide distributions in the solid or liquid phase in water–rock geochemical systems. Accordingly, this study reviewed previous research on the applications of geochemical models, including studies on the development of geochemical models, sources of thermodynamic databases (TDBs) and their applications in programs, the determination of the adequacy of TDBs in surface complexation models and case studies, and the selection and application of activity coefficient equations in geochemical models. In addition, the study conducted case studies and comparisons of the activity coefficients derived by different geochemical models. Three activity coefficient equations, namely the Davies, modified Debye–Hückel, and Pitzer equations, and four geochemical models, namely PHREEQC, MINEQL+, MINTEQA2, and EQ3/6, were used in the study. The results demonstrated that when the solution’s ionic strength was <0.5 m, the differences in the activity coefficients between the Davies and modified Debye–Hückel equations were <5%. The difference between the Pitzer and Davies equations, or between the Pitzer and modified Debye–Hückel equations in terms of the calculated activity coefficients was <8%. The effect of temperature on the activity coefficient slightly influenced the modeling outputs of the Davies and modified Debye–Hückel equations. In the future, the probability distribution and uncertainty of parameters of Kd and the equilibrium constant can be used in geochemical and reactive transport models to simulate the long-term safety of nuclear waste disposal sites. The findings of this study can provide a strong scientific basis for conducting safety assessments of nuclear waste disposal repositories and developing environmental management or remediation schemes to control sites marred by near-surface contamination.

Natural and Anthropogenic Analogues for High-Level Nuclear Waste Disposal Repositories: A Review

ABSTRACTProjects involving deep geological disposal of nuclear waste are unique in that predictive models of long-term safety often involve geological timeframes. This manuscript provides a review of natural and anthropogenic analogues for high-level nuclear waste disposal in a deep geological repository. We also occasionally highlight analogues that have been used for low- and intermediate-level waste. Most studies define natural analogues as either naturally occurring or anthropogenic systems. In this paper, we distinguish between natural analogues and anthropogenic analogues because the latter generally provide non-technical (anecdotal) illustrations of concepts and processes for the safety case, whereas the former can provide technical and quantitative information. In addition, natural analogues can provide information over geological time scales (millions of years) and spatial scales (kilometers), whereas anthropogenic analogues provide information over a much more limited time scale (hundreds or thousands of years). Regardless of the definition, analogue studies provide one of the multiple lines of evidence intended to increase confidence in the safe geological disposal of high-level nuclear waste. They are deemed necessary because they complement the experiments that are carried out over a period of months or years. They also provide a way to validate numerical long-term safety assessment models with information and data covering geological time scales and spatial scales.The first part of this review describes the analogue concept. The second and third parts provide examples of natural and anthropogenic analogues for engineered barrier systems and natural barriers. Part four describes analogues for complex coupled transport processes. Finally, we discuss general and specific areas of future research.A concerted effort should be made to ensure that there is a transfer of data from the complex, natural analogue field studies to simplistic models which, by necessity, are used to evaluate the long-term safety of deep geological repositories. Field analogue studies should be planned to interface with laboratory experiments and, ultimately, with in situ field experiments, when the final repository site is selected. This will involve using natural analogue data in a quantitative way to support the deep geological repository safety case.

Transport behaviors of Cs+ in granite porous media: Effects of mineral composition, HA, and coexisting cations

The transport of radiocesium (RCs) in granite has attracted great concerns for the consideration of a long-term safety assessment and performance evaluation of the nuclear waste disposal repository. In this study, the transport behaviors of Cs+ in granite were addressed and quantified by column experiments, sequential extraction, and a convection-dispersion equation model. The transport of Cs+ in granite experienced at least two stages including a rapid increase and a slow increase stages. The retardation of Cs+ in granite obviously became higher as biotite content increased. However, a consistent breakthrough plateau and almost overlapped breakthrough curves were observed under different feldspar contents, which suggested that the transport behaviors of Cs+ in granite was quite close to feldspar. Compared to Na+, K+ could effectively inhibit Cs+ adsorption and facilitate the mobility of Cs+ in granite column. In the presence of Sr2+, the transport of Cs+ was provoked in the granite column mainly due to the high competition effects. Humic acid (HA) did not obviously change the transport behaviors of Cs+ in granite column; however, HA could weakly change the adsorption species of Cs+ during Cs+ transport in granitic media. Both sequential extraction and two-site non-equilibrium model suggested that feldspar was the main contributor to the weak adsorption sites and biotite was responsible for the strong affinity sites for Cs+ in Beishan granite. The findings could provide important insights into RCs transport and fate in granitic media.

A Thermo-Hydro-Mechanical Coupling Analysis for the Contaminant Transport in a Bentonite Barrier with Variable Saturation

The migration of hazardous substances in a bentonite barrier layer is a key issue for the safety of nuclear waste storage. This study develops a thermo-hydro-mechanical coupling model to simulate the contaminant migration in a bentonite barrier layer of the nuclear waste storage chamber. In particular, the Richard’s equation is used to describe the groundwater flow in the bentonite barrier layer with variable saturation. Thermal diffusion and concentration diffusion are coupled with the layer deformation and fluid flow. The migration rate and diffusion range of hazardous substances in the bentonite barrier layer are numerically simulated. These numerical simulations show that the heat release from nuclear waste can induce a temperature gradient and deformation, and thus significantly affects both transfer rate and concentration distribution of dilute substances. These multi-physical couplings under different initial saturation may significantly modify the sealing efficiency of an unsaturated bentonite barrier layer and thus this model is of great significance in the safety evaluation of a nuclear waste disposal repository.

Permeability evolution and crack characteristics in granite under treatment at high temperature

The evolution of permeability and microcrack characteristics of granite at high temperature has significant effect on the safe and stable operation of high-level nuclear waste disposal repositories. We measure the permeability, porosity and formation factors of granite specimens following thermal treatment including the use of NMR to certify observed response. The results indicate that the initial and residual permeability and porosity change little when T ≤ 300 °C, but increase rapidly when 300 °C ≤ T ≤ 600 °C before entering a stable phase when 600 °C ≤ T ≤ 750 °C. At T = 150 °C microcrack apertures and radius only slight increase while microcrack density and fraction of connectedness all slight decrease, but net causing porosity, the peak strength and elastic modulus to increase. The strength and elastic modulus decrease exponentially with fractional connectivity of microcrack while the initial compressibility of microcracks scale linearly with initial aperture. The formation factor for the granite increases near-linearly with effective stress with the rate of increase generally decreasing with increasing temperature. Thermal conductivity first increases rapidly before remaining constant with increasing effective stress. With increasing temperature, the thermal conductivity decreases and becomes more sensitive to the effective stress.

Analysis of permeability in rock fracture with effective stress at deep depth

In this study, the application of conventional cubic law to a deep depth condition was experimentally evaluated. Moreover, a modified equation for estimating the rock permeability at a deep depth was suggested using precise hydraulic tests and an effect analysis according to the vertical stress, pore water pressure and fracture roughness. The experimental apparatus which enabled the generation of high pore water pressure (< 10 MPa) and vertical stress (< 20 MPa) was manufactured, and the surface roughness of a cylindrical rock sample was quantitatively analyzed by means of 3D (three-dimensional) laser scanning. Experimental data of the injected pore water pressure and outflow rate obtained through the hydraulic test were applied to the cubic law equation, which was used to estimate the permeability of rock fracture. The rock permeability was estimated under various pressure (vertical stress and pore water pressure) and geometry (roughness) conditions. Finally, an empirical formula was proposed by considering nonlinear flow behavior; the formula can be applied to evaluations of changes of rock permeability levels in deep underground facility such as nuclear waste disposal repository with high vertical stress and pore water pressure levels.

A numerical study on the coupled hydro-mechanical behaviour of compacted bentonite

Understanding the mechanical behaviour of compacted bentonite upon re-saturation is of outmost importance in most designs of nuclear waste disposal repositories. The behaviour of bentonite is characterized by its stress-path dependency and it is typically interpreted on the basis of its microstructural interactions. Up to now, effective stress-based models have had limited success in reproducing consistently the main responses. Here a recently proposed model for the modelling of volumetric behaviour of compacted bentonites is extended to triaxial stress states. The model is formulated using a conventional effective stress expression and the degree of saturation. Because these two variables are directly related to the water retention, a suitable formulation for bentonites is used. The resulting equations are characterised by a high degree of hydro- mechanical couplings and a low number of material parameters, which can be obtained on the basis of well- established laboratory procedures. In order to demonstrate its capabilities, the model is used to simulate the behaviour of MX-80 bentonite for several stress paths under oedometric conditions. The emphasis is put on the process of parameter determination. The predictive capabilities of the model are also highlighted.

Swelling pressure of Na- and Ca-montmorillonites in saline environments: A molecular dynamics study

The swelling pressure of bentonite clay is an important characteristic that influences its use in nuclear waste disposal repositories. A molecular dynamics study was performed to investigate the influence of salt solutions of different concentrations on the swelling pressure of Na-montmorillonite and Ca-montmorillonite in NaCl and CaCl2 solutions, respectively. Na-montmorillonite shows a gradual increase in swelling pressure as the dry density increases, while Ca-montmorillonite shows a slow increase as the dry density increases, after which a steeper increase in swelling pressure was observed. There is an overall decrease in swelling pressure of Na- and Ca-montmorillonites as the salinity of the surrounding solution increases. The influence of ionic strength was most prominent for Na-montmorillonite at moderate to high dry densities. At low to moderate dry densities, the swelling pressure of Na-montmorillonite was greater than that of Ca-montmorillonite; however, at high dry densities and ionic strengths, Ca-montmorillonite exhibits the higher swelling pressures.

Solubility of U(VI) in chloride solutions. III. The stable oxides/hydroxides in MgCl2 systems: Pitzer activity model for the system UO22+–Na+–K+–Mg2+–H+–OH−–Cl−–H2O(l)

We have developed new chemical, thermodynamic and activity models for the system UO22+–Na+–K+–Mg2+–H+–OH−–Cl−–H2O(l) within the Pitzer approach. The new thermodynamic model is based on previously reported data treated within the SIT approach for NaCl and KCl systems, as well as on new experimental data determined in this work for the MgCl2 system.The solubility of uranium(VI) was studied in 0.01–5.15 mol·kgw−1 MgCl2 solutions at pHm = 4.1–9.7 (with pHm = −log [H+]). Experiments were performed under Ar atmosphere at T = (22 ± 2) °C. In all investigated systems, the solubility of U(VI) is controlled by metaschoepite, UO3·2H2O(cr). In contrast to previously investigated NaCl and KCl systems, no ternary Mg–U(VI)–OH(s) solid phases formed in alkaline MgCl2 solutions within the timeframe of this study (≤ 200 days). A very significant increase in the solubility (up to 3 log10-units) is observed in acidic to near-neutral pHm conditions when increasing MgCl2 concentration from 0.01 to 5.15 mol·kgw−1, which reflects the strong ion interaction processes taking place in concentrated MgCl2 brines.The solubility of UO3·2H2O(cr) in the investigated NaCl, KCl and MgCl2 solutions is well described with the solubility and hydrolysis constants recommended by Altmaier et al., (2017) and NEA–TDB, and a Pitzer activity model derived in the present work. The latter model considers experimental data reported in the present study and available in the literature for NaCl, KCl and MgCl2 systems (solubility, potentiometric and spectroscopic data), in combination with well-stablished estimation methods and correlations with SIT coefficients.Chemical, thermodynamic and Pitzer activity models provided in this work for the system UO22+–Na+–K+–Mg2+–H+–OH−–Cl−–H2O(l) accurately describe all evaluated datasets, and represent an adequate tool for the calculation of U(VI) solubility and aqueous speciation in a variety of geochemical conditions including concentrated brine systems of relevance in salt-based repositories for nuclear waste disposal.

Enthalpies of formation of polyhalite: A mineral relevant to salt repository

Polyhalite is an important coexisting mineral with halite in salt repositories for nuclear waste disposal, such as Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. The thermal stability of this mineral is a key knowledge in evaluating the integrity of a salt repository in the long term, as water may release due to thermal decomposition of polyhalite. Previous studies on structural evolution of polyhalite at elevated temperatures laid the basis for detailed calorimetric measurements. Using high-temperature oxide-melt drop-solution calorimetry at 975K with sodium molybdate as the solvent, we have determined the standard enthalpies of formation from constituent sulfates (ΔH°f,sul), oxides (ΔH°f,ox) and elements (ΔH°f,ele) of a polyhalite sample with the composition of K2Ca2Mg(SO4)4·1.95H2O from the Salado formation at the WIPP site. The obtained results are: ΔH°f,sul=−152.5±5.3kJ/mol, ΔH°f,ox=−1926.1±10.5kJ/mol, and ΔH°f,ele=−6301.2±9.9kJ/mol. Furthermore, based on the estimated formation entropies of polyhalite, its standard Gibbs free energy of formation has been derived to be in the range of −5715.3±9.9kJ/mol to −5739.3±9.9kJ/mol. These determined thermodynamic properties provide fundamental parameters for modeling the stability behavior of polyhalite in salt repositories.

Studies on corrosion behaviour of low carbon steel canister with and without γ-irradiation in China’s HLW disposal repository

ABSTRACTCarbon steel is the candidate material of the canister in high level radioactive nuclear waste disposal repository in China. The corrosion behaviour of X65 grade low carbon steel exposed to bentonite thermally aged at 363.15 K after exposure to 0 and 2.98 kGy h−1 γ-irradiation for 1007 h was studied using optical microscopy, scanning electron microscopy, energy dispersive X-ray spectroscopy, X-ray diffraction, micro-Raman spectroscopy (μ-Raman), weight loss measurements and white-light interferometry. The results show that 2.98 kGy h−1 γ-irradiation causes the formation of new phases of siderite and maghemite, and accelerates the corrosion rate by ∼33%.This paper is part of a supplement on the 6th International Workshop on Long-Term Prediction of Corrosion Damage in Nuclear Waste Systems.

Effect of sample size on the fluid flow through a single fractured granitoid

Most of deep geological engineered structures, such as rock caverns, nuclear waste disposal repositories, metro rail tunnels, multi-layer underground parking, are constructed within hard crystalline rocks because of their high quality and low matrix permeability. In such rocks, fluid flows mainly through fractures. Quantification of fractures along with the behavior of the fluid flow through them, at different scales, becomes quite important. Earlier studies have revealed the influence of sample size on the confining stress–permeability relationship and it has been demonstrated that permeability of the fractured rock mass decreases with an increase in sample size. However, most of the researchers have employed numerical simulations to model fluid flow through the fracture/fracture network, or laboratory investigations on intact rock samples with diameter ranging between 38 mm and 45 cm and the diameter-to-length ratio of 1:2 using different experimental methods. Also, the confining stress, σ3, has been considered to be less than 30 MPa and the effect of fracture roughness has been ignored. In the present study, an extension of the previous studies on “laboratory simulation of flow through single fractured granite” was conducted, in which consistent fluid flow experiments were performed on cylindrical samples of granitoids of two different sizes (38 mm and 54 mm in diameters), containing a “rough walled single fracture”. These experiments were performed under varied confining pressure (σ3 = 5–40 MPa), fluid pressure (fp ≤ 25 MPa), and fracture roughness. The results indicate that a nonlinear relationship exists between the discharge, Q, and the effective confining pressure, σeff., and Q decreases with an increase in σeff.. Also, the effects of sample size and fracture roughness do not persist when σeff. ≥ 20 MPa. It is expected that such a study will be quite useful in correlating and extrapolating the laboratory scale investigations to in-situ scale and further improving theoretical/numerical models associated with fluid flow through rock masses.

Tc and Re behavior in borosilicate waste glass vapor hydration tests II

Technetium (99Tc) is a significant environmental risk factor to consider for nuclear waste disposal repositories. Rhenium (Re), in the same column of the periodic table as Tc, is often used as a non-radioactive surrogate for Tc. Six waste glasses containing both Tc and Re were synthesized under a variety of redox conditions to produce different distributions of Tc and Re oxidation states. These glasses were exposed to vapor hydration tests (VHT) at 200°C for 23 to 30days; and the Tc and Re oxidation state, coordination environment, and spatial distribution within the altered coupons were determined. Compared with the original glasses, the corresponding VHT samples showed substantial reduction of Tc species, except where the original glass contained only reduced Tc (Tc4+). Similar to earlier findings, Tc is more sensitive to redox conditions than Re with respect to both glass synthesis conditions and VHT alteration processes. Glasses that originally contained more oxidized Tc (near 100% Tc7+) showed the most Tc enrichment in the altered VHT sample layers, where Tc was largely reduced to Tc4+. Re is generally more oxidized than Tc in the samples measured and has similar spatial distributions as Tc in some VHT samples, while having very different spatial distributions compared with Tc in others. Glasses that originally had a distribution of Tc oxidation states (approximately 1:1 Tc4+ to Tc7+), had Tc concentrations in the VHT altered layers that were approximately equal to or less than those found in the unaltered glass. However, in the same samples, Re concentrations were highest in the altered layers. Overall, with regard to spatial distributions within the altered VHT layers, the behavior of Re was not a good predictor of Tc behavior. Therefore, at least under VHT conditions, using Re as a non-radioactive surrogate for Tc in borosilicate waste glasses can provide misleading results.

The coupled thermal-hydraulic-mechanical behaviour of a large-scale in situ heating experiment

The thermo-hydraulic-mechanical behaviour of a large in situ heating test, the Buffer/Container Experiment (BCE), carried out at Atomic Energy of Canada's underground research laboratory, is considered. The work can be seen as an extension of the authors' previous research, investigating the behaviour of a large-scale test, the so-called ‘isothermal test’, to now include heating effects. Suggestions related to the micro–macro behaviour of bentonitic buffer materials subjected to resaturation in confined non-isothermal conditions are explored. Simulation of pre-heating phases of the experiment demonstrates the ability of the model to describe the hydraulic regime in the host rock and isothermal infiltration into the buffer. Consideration of the heating phase confirms that the temperature field is well understood, with a good correlation between the numerical and experimental results. In the moisture field it is found that the interaction between the emplaced buffer and host rock is of considerable importance. It is found that inclusion of micro–macro interaction effects, by way of consideration of the impact of swelling and micro–macro interactions on the hydraulic conductivity of the material, yielded significantly improved correlations between observed and calculated results in the moisture field. The results of the numerical simulations suggest that the interaction between the emplaced buffer and host rock is strongly influenced by the micro–macro behaviour of the bentonitic buffer material. It is concluded that the consideration of the impact of micro–macro interactions on moisture flow in bentonitic buffers is of considerable importance, and may be of some significance in considering the total resaturation time as part of the performance assessment of a nuclear waste disposal repository.