High-level Waste Repository Research Articles

Solubility study of Tc(IV) in a granitic water

The deep geological disposal of the high level radioactive wastes is expected to be a safe disposal method in most countries. The long-lived fission product99Tc is present in large quantities in nuclear wastes and its chemical behavior in aqueous solution is of considerable interest. Under oxidizing conditions technetium exists as the anionic species TcO4−whereas under the reducing conditions, expected to exist in a deep geological repository, it is generally predicted that technetium will be present as TcO2·nH2O. Hence, the mobility of Tc(IV) in reducing groundwater may be limited by the solubility of TcO2·nH2O under these conditions. Due to this fact it is important to investigate the solubility of TcO2·nH2O. The solubility determines the release of radionuclides from waste form and is used as a source term in radionuclide migration analysis in performance assessment of radioactive waste repository. Technetium(IV) was prepared by reduction of a technetate solution with Sn2+. The solubility of Tc(IV) has been determined in simulated groundwater and redistilled water under aerobic and anaerobic conditions. The effects of pH and CO32−concentration of solution on solubility of Tc(IV) were studied. The concentration of total technetium and Tc(IV) species in the solutions were periodically determined by separating the oxidized and reduced technetium species using a solvent extraction procedure and counting the beta activity of the99Tc with a liquid scintillation counter. The experimental results show that the rate of oxidation of Tc(IV) in simulated groundwater and redistilled water is about (1.49∼1.86)×10−9mol L−1d−1under aerobic conditions, while no Tc(IV) oxidation was detected in simulated groundwater and redistilled water under anaerobic conditions. Under aerobic or anaerobic conditions the solubility of Tc(IV) in simulated groundwater and redistilled water is equal on the whole after centrifugation or ultrafiltration. The solubility of Tc(IV) increases with the decrease of pH at pH<2, increases with the increase of pH at pH>11 and is pH independent in the range 2<pH<11. The concentrations of Tc(IV) species were in the range of 10−8to 10−9mol L−1at 2<pH<11. The solubility of Tc(IV) slightly increases with increasing the increase of CO32−concentration. Geochemical modelling showed a good agreement between our experimental results and thermodynamic constants from the NEA TDB review. These data could be used to estimate the Tc(IV) solubility for cases where solubility limits transport of technetium in reducing environments of high-level waste repositories.

Estimation of loading density of underground well repositories for solid high-level radioactive wastes

The convective transfer of radionuclides by subsurface water from a geological repository of solidified high-level radioactive wastes (HLW) is considered. The repository is a cluster of wells of large diameter with HLW disposed of in the lower portions of the wells. The safe distance between wells as a function of rock properties and parameters of well loading with wastes has been estimated from mathematical modeling. A maximum permissible concentration of radionuclides in subsurface water near the ground surface above the repository is regarded as a necessary condition of safety. The estimates obtained show that well repositories allow for a higher density of solid HLW disposal than shaft storage facilities. Advantages and disadvantages of both types of storage facilities are considered in order to estimate the prospects for their use for underground disposal of solid HLW.

Hydro-mechanical, geochemical and mineralogical characteristics of the bentonite buffer in a heater experiment: The HE-B project at the Mont Terri Rock Laboratory

A small scale heater experiment was realized at the underground laboratory in Mont Terri, Switzerland. The effects of heat-producing waste in a repository for high-level waste and spent fuel were simulated by a heater element of 10 cm diameter, held at a constant surface temperature of 100 °C. The heater element was embedded in an engineered barrier of compacted bentonite blocks that were water saturated for 35 months before the 18 months heating phase was started. The host rock is a highly consolidated stiff Jurassic clay stone (Opalinus Clay). Extensive on-site and laboratory tests were performed to determine the hydro-mechanical, geochemical and mineralogical characteristics of the bentonite buffer to identify variations in the material properties in response to heating. Other studies deal with the host rock material. The results indicate that the bentonite shows only very weak modifications during the heater experiment. Whereas mineralogical alterations (e.g. transformations and formation of new phases) could not be measured, very weak changes in layer charge density and interlayer chemical composition of the smectite were found. Changes in the porosity and surface area of the bentonite may be explained as result of cementing or aggregation processes. These processes affect some THM characteristics like porosity, water uptake and thermal properties.

WARREN K. SINCLAIR KEYNOTE ADDRESS: CONTEMPORARY ISSUES IN RISK-INFORMED DECISION MAKING ON THE DISPOSITION OF RADIOACTIVE WASTE

A consistent and transparent risk-informed approach to managing nuclear waste is plagued with different regulators, different rules and regulations for different waste types, different compliance requirements, and indecisions about probabilistic vs. deterministic models. Low-activity waste management is particularly void of a path forward with respect to being risk-informed. Risk assessment is not referenced in the statutes on low-activity waste even though both the U.S. Environmental Protection Agency and U.S. Nuclear Regulatory Commission (U.S. NRC) have policies to apply consistent risk management approaches to all of their programs. The U.S. NRC has developed guidance on the preparation of probabilistic performance assessments for low-activity waste facilities, but there have been no serious takers and a lack of initiative on the part of licensees. Thus, little to no experience exists on risk-informing low-activity waste. The missed opportunities include establishing a risk basis that would allow for simpler, safer, and much less costly alternatives for low-activity waste disposal while enabling society to have the full benefit of radiation technologies. There is hope that congressional action or regulatory rule making will address some of these issues with the result being the adoption of a more general and unified approach to risk-informed regulation of all types of waste. Just as much of the initiative for risk-informed nuclear power came from industry, it must also be the case for nuclear waste. A start would be the adoption of a basic framework of risk assessment in waste management applicable to all types of waste--radioactive and nonradioactive. The "set of triplets" risk assessment framework that is applicable to any kind of risk is an established alternative. It is believed that such a framework with the support of a regulatory structure made compatible through appropriate rulemaking or congressional action, and the experience of the probabilistic performance assessments for the Waste Isolation Pilot Plant and the proposed Yucca Mountain high-level waste repository, could result in the right path forward for the regulation and management of low-activity waste.

Self-sealing barriers of clay/mineral mixtures – The SB project at the Mont Terri Rock Laboratory

Moderately compacted clay/mineral mixtures may represent a reasonable alternative to highly compacted bentonite buffers currently studied and considered in some concepts of underground repositories for high-level radioactive wastes. In contrast to highly compacted buffers clay/sand mixtures exhibit a high permeability to gas in the unsaturated state and a comparably low gas entry/break through pressure in the saturated state while providing an adequate self-sealing potential due to swelling of the clay minerals after water uptake from the host rock. By using optimized material mixtures, the evolution of high gas pressure in the repository near-field due to corrosion of the waste containers will be avoided and possible migration of radionuclides from the waste matrix in the liquid phase through the buffer will be diffusion controlled just like in the host rock. On basis of promising laboratory results gained in GRS’ geotechnical laboratory it was decided to test and demonstrate the sealing properties of clay/mineral mixtures under realistic in situ conditions at the Mont Terri Underground Rock Laboratory (MTRL). The paper presents details about the envisaged in situ experiments and material data obtained from laboratory investigations. First results of full-scale mock-up tests are presented as well. In addition, information is given about further laboratory investigations and scoping calculations that have been performed to analyze whether it would be possible to achieve and demonstrate the required sealing properties within the comparably short run time of the project. It has been found that clay/sand mixtures with clay contents between 35% and 50% are suitable for the envisaged in situ tests at the MTRL (and most likely also for adequate sealing of disposal rooms in repositories). The SB project is part of the Integrated Project ESDRED [ANDRA, 2005. www.esred.info] funded by the Commission of the European Commission.

Performance of the bentonite barrier at temperatures beyond 100 °C: A critical review

Many current concepts for the engineered barriers of high-level waste repositories include a bentonite buffer or backfill which will be exposed to significantly raised temperatures for a period of decades to centuries. To assess its stability, available experimental and natural analogue data on thermally-exposed bentonite is reviewed. The two main alteration processes identified are cementation by precipitation of SiO 2 and transformation of smectite to non-expandable illite layers. These may lead to reduced swelling and loss of plasticity as well as increased hydraulic conductivity and diffusivity. Relevant data for repository conditions from both laboratory and field studies is scarce, but nevertheless yields a fairly consistent picture. Thus, no significant changes of hydraulic and mechanical properties have been reported for bentonite materials exposed to temperatures of at least 120 °C under wet conditions. The data suggest significant cementation and perhaps also illitisation effects occur at 150 °C and beyond. Interestingly, natural analogue bentonite samples that showed substantial cementation and illitisation effects still displayed rather favourable hydraulic properties. Under dry conditions, bentonite is stable to higher temperatures – maybe as high as 350 °C. In order to verify the findings from this work and to get more reliable information at temperatures beyond 130 °C, a series of long-term experiments examining hydraulic, mechanical and mineralogical changes under realistic conditions would be useful.

Deliquescence behavior of salts deposited inside the drifts of a potential high-level waste repository

Deliquescence behavior of salts deposited inside the drifts of a potential high-level waste repository

Application of controlled blasting during the construction of the Canadian underground research laboratory

The Korean Atomic Energy Research Institute is currently planning the construction of an Underground Research Laboratory to carry out research and development related to the disposal of high-level wastes from nuclear reactors used to generate electrical power. This paper discusses the excavation methods used to construct the Canadian Underground Research Laboratory and their application in planning for the construction of a similar underground laboratory and eventually an underground repository for high-level wastes in Korea. (A) This paper was presented at Safety in the underground space - Proceedings of the ITA-AITES 2006 World Tunnel Congress and the 32nd ITA General Assembly, Seoul, Korea, 22-27 April 2006. For the covering abstract see ITRD E129148. Reprinted with permission from Elsevier.

Microbiological effects on the Cavern Extended Storage (CES) repository for radioactive waste—A quantitative evaluation

The recently developed Cavern Extended Storage (CES) repository design for radioactive waste combines some of the advantages of the flexibility provided by long-term interim storage with the safety of deep geological disposal. It is also of particular interest where demands are made for long periods of institutional control and monitoring (or staged implementation) of repositories or where a desire exists to keep options open for possible retrieval with regard to spent fuel reprocessing. In this concept, large steel casks containing vitrified High Level Waste (HLW) and/or spent fuel are stored in large rock caverns at depths 500 to 1000 m below surface for approximately 300 years before backfilling to transform this ‘store’ into a final ‘repository’. However, the extreme environmental conditions of a HLW repository cannot guarantee sterility. As a result, the humid, ventilated caverns could provide many potential niches for microbial growth, which could influence both the integrity of the storage casks (and any associated infrastructure) and the subsequent long-term performance of the sealed repository. The storage period is likely to provide the optimum period for microbial colonisation and growth, with the casks and rock walls exposed to constantly circulating air, which would provide oxygen for aerobic microbial processes. In addition, nutrients and energy sources will be introduced from the surface during operational and monitoring procedures. The duration of this storage period may be extended or shortened, depending on changing institutional requirements and it is thus particularly important to know the effects of microbial activity during this time. Simple models of the constraints on such growth can be used to scope the potential for activity to influence repository performance, particularly in relation to changing storage times, and to compare the CES design with conventional repository concepts.

Managing fusion high-level waste—A strategy for burning the long-lived products in fusion devices

Fusion devices appear to be a viable option for burning their own high-level waste (HLW). We propose a novel strategy to eliminate (or minimize) the HLW generated by fusion systems. The main source of the fusion HLW includes the structural and recycled materials, refractory metals, and liquid breeders. The basic idea involves recycling and reprocessing the waste, separating the long-lived radionuclides from the bulk low-level waste, and irradiating the limited amount of HLW in a specially designed module to transmute the long-lived products into short-lived radioisotopes or preferably, stable elements. The potential performance of the new concept seems promising. Our analysis indicated moderate to excellent transmutation rates could be achieved in advanced fusion designs. Successive irradiation should burn the majority of the HLW. The figures of merit for the concept relate to the HLW burn-up fraction, neutron economy, and impact on tritium breeding. Hopefully, the added design requirements could be accommodated easily in fusion power plants and the cost of the proposed system would be much less than disposal in a deep geological HLW repository. Overall, this innovative approach offers benefits to fusion systems and helps earn public acceptance for fusion as a HLW-free source of clean nuclear energy.

Evolution of Chemistry and Its Effects on the Corrosion of Engineered Barrier Materials

ABSTRACTThe evolution of environmental conditions within the emplacement drifts of a potential high-level waste repository at Yucca Mountain, Nevada, may be influenced by several factors, including the temperature and relative humidity within the emplacement drifts and the composition of seepage water. The performance of the waste package and the drip shield may be affected by the evolution of the environmental conditions within the emplacement drifts. In this study, tests evaluated the evolution of environmental conditions on the waste package surfaces and in the surrounding host rock. The tests were designed to (i) simulate the conditions expected within the emplacement drifts; (ii) measure the changes in near-field chemistry; and (iii) determine environmental influence on the performance of the engineered barrier materials. Results of tests conducted in this study indicate the composition of salt deposits was consistent with the initial dilute water chemistry. Salts and possibly concentrated calcium chloride brines may be more aggressive than either neutral or alkaline brines.

“Perspective The origin of the EPA' s 10 000-year time frame for the high-level waste repository” A. Alan Moghissi

“Perspective The origin of the EPA' s 10 000-year time frame for the high-level waste repository” A. Alan Moghissi

Powder diffraction of sodalite in a multiphase ceramic used to immobilize radioactive waste

The title compound, ∣Na6Li1.6K0.4Cl2∣[Al6Si6O24]‐SOD, is similar to sodalite proper, but the introduction of Li and K into the structure creates a reduction in unit-cell volume and additional collapse of the framework tetrahedra. Refinement of an X-ray powder diffraction pattern of a multiphase material yielded for sodalite a lattice parameter of 0.88427 (2) nm, an Al–O–Si bond angle of 137.9(3°), and Al–O and Si–O bond lengths of 0.1730(5) nm and 0.1620(5) nm, respectively. The angle of the unique Al–O–Si bond corresponds well with the 138° obtained by 29Si solid-state magic-angle-spinning nuclear magnetic resonance spectroscopy. This characterization is important since the compound constitutes an essential part of a radioactive waste form intended for a high-level waste repository.

Tensit - A simulation tool for nuclide transport, risk and dose calculations

The present paper presents, a probabilistic simulation package, Tensit, used for migration calculation of radionuclides from a deep repository for spent nuclear fuel to the biosphere. The package is based on pre-existing codes (Matlab, Simulink and the probabilistic engine @Risk) and is capable of performing migration calculations for the near field, far field and the biosphere for a high level waste repository. In addition to the compartment based models based on transfer factors, the code is also able to handle the more complex ecosystem models (based on flow of carbon and nutrients) either separately or in combination with the compartment models. Comparisons with other codes used for compartment based biosphere modelling as well as example of simple models, more advanced ecosystems and connected ecosystems Landscape model are shown in the paper.

Performance assessment model development and parameter acquisition for analysis of the transport of natural radionuclides in a Mediterranean watershed

This paper describes the methodology developed to construct a model for predicting the behaviour of the natural radioisotopes of U, Th and Ra in a Mediterranean watershed. The methodology includes the development of the performance assessment model, obtaining water flow and radiological parameters based on experimental data and analysis of results. The model, which accounts for both water flows and mass balances of the radionuclides in a semi-natural environment, provides assessments of radionuclide behaviour in grassland and agricultural soils, rivers and reservoirs, including the processes of radionuclide migration through land and water and interactions between both. From field and laboratory data, it has been possible to obtain parameters for the driving processes considered in the model, water fluxes, source term definition, soil to plant transfer factors and distribution coefficient values. Ranges of parameter values obtained have shown good agreement with published literature data. This general methodological approach was developed to be extended to other radionuclides for the modelling of a biosphere watershed in the context of performance assessment of a High Level Waste (HLW) repository under Mediterranean climate conditions, as well as for forecasting radionuclide transport under similar Mediterranean conditions that will occur in the future in other areas. The application of sensitivity and uncertainty analysis was intended to identify key uncertainties with the aim of setting priorities for future research. The model results for the activity concentration in the reservoir indicate that for 238U and 230Th the most relevant parameter is the initial concentrations of the radionuclides in the reservoir sediments. However, for 226Ra the most important parameter is the precipitation rate over the whole watershed.

Surface charges and Np(V) sorption on amorphous Al and Fe silicates

Sorption onto Si-rich alteration layers of crystalline minerals and nuclear glasses, and onto amorphous secondary silicates of rocks and soils, are expected to retard the migration of actinides in the near- and far-field of high-level waste repositories. In this work, we present experimental and modeling studies on the effects of silicate structure and bulk chemistry, and of solution chemistry, on charges and adsorption of neptunyl ions at surfaces of synthetic, amorphous or poorly ordered silica, Al silicates and Fe silicates. The Al silicates display similar pH-dependent surface charges characterized by predominant Si–O − sites, and similar surface affinities for neptunyl ions, irrespective to their Si/Al molar ratio (varying from 10 to 4.3). Such experimental features are explained by incorporation of Al atoms in tetrahedral position in the silicate lattice, leading to only trace amounts of high-affinity Al–OH surface groups due to octahedral Al. By contrast, the structure of the Fe silicates ensures the occurrence of high-affinity Fe–OH surface groups, whose concentration is shown by proton adsorption measurements to increase with decreasing of the silicate Si/Fe molar ratio (from 10 to 2.3). Nevertheless, experimental data of the adsorption of neptunyl and electrolyte ions show unexpectedly weak effect of the Si/Fe ratio, and suggest predominant Si–OH surface groups. A possible explanation is that aqueous silicate anions, released by dissolution, adsorb at Fe–OH surface groups and/or precipitate as silica gel coatings, because experimental solutions were found at near-equilibrium with respect to amorphous silica. Therefore, the environmental sorption of Np(V) onto Si-rich, amorphous or poorly ordered Al silicates may primarily depend on pH and silicate-specific surface areas, given the low overall chemical affinity of such phases for dissolved metals. By contrast, the sorption of Np(V) on natural, amorphous or poorly ordered Fe silicates may be a complex function of silicate bulk chemistry and solution chemistry, i.e., of pH and aqueous Si concentrations. Simple conceptual models of the surface chemistry of the Al and Fe silicates are developed here, based on the wealth of experimental data of silicate surface charges. The surface complexation models predict reasonably the effect of solution chemistry on the sorption of neptunyl ions on poorly ordered silicates of various compositions, and can thus be useful in extrapolating neptunyl mobility in many geochemical systems.

Modelling bentonite–water interactions at high solid/liquid ratios: swelling and diffuse double layer effects

Previous experimental studies on bentonite–water interactions have demonstrated the importance of ion exchange and surface complexation reactions occurring at the clay surface as well as dissolution/precipitation reactions of mineral impurities. Based on these findings, thermodynamic models have been widely used to derive porewater compositions for the compacted bentonite used as backfill material for nuclear repositories. Conventional models typically neglect phenomena important in compacted clays such as anion exclusion induced by swelling of the expandable clay fraction and by the formation of electrical double layers on charged edge surfaces. In this study, we evaluate such phenomena by applying a refined diffuse double layer (DDL) approach to model porewater composition in a compacted bentonite backfill surrounded by argillaceous host rock, as foreseen for the Swiss high-level waste repository. Model calculations also include the effect of water incorporation in the structural interlayers. The results indicate that the conventional model and the refined DDL model without distinction between interlayer and external water only differ slightly. The main buffering reactions include ion exchange of Ca for Na, calcite and gypsum dissolution and deprotonation of surface hydroxyl groups. On the other hand, the calculation accounting for the distinction of external and interlayer water indicates significant anion exclusion effects on the external water composition. Most notably, this leads to an increased salinity and drop in pH. From a performance assessment perspective, however, the differences induced by the inclusion of swelling and diffuse double layer effects are not very significant relative to uncertainties related to system variables, such as the pCO 2 of the host rock. Finally, it should be emphasised that significant uncertainties related to the thermodynamic properties of water in compacted clays, e.g., dielectric constant, are still unresolved and deserve further investigations.

The geochemical behaviour of Tc, Np and Pu in spent nuclear fuel in an oxidizing environment

Abstract Spent fuel from commercial nuclear reactors consists mainly of uranium oxide. However, the changes that occur during reactor operations have a profound effect on chemical and physical properties of this material. Heat build-up in the fuel pellet during reactor operations can cause redistribution of fission products. The fission products may aggregate in one of three types of precipitates; gaseous, metallic, or oxide, depending on the burn-up and in-core treatment. Radiation damage and variations in fission and neutron capture yields across the fuel pellets lead to Pu enrichment and increased porosity with increasing burn-up. A more porous surface may make the fuel more susceptible to oxidative dissolution. As the level of actinides and fission products increases, the fuel may become more resistant to oxidation. These changes may limit the usefulness of natural uraninite (UO 2 ) analogues for predicting the geological behaviour of spent fuel disposed in a high-level waste (HLW) repository. In this Chapter, an overview of spent fuel microstructure, radiolytic effects, and alteration processes is presented. Evidence for Np incorporation into U 6+ phases, the nature of Pu surface precipitates on spent fuel, and evidence for the preferential removal of 4d-metals from ε-particles in corroded spent fuel is discussed. Understanding the potential mechanisms of radionuclide attenuation through sorption and/or incorporation requires techniques with both high spatial resolution and excellent elemental sensitivity.

Comparative Fuel Cycle Analysis of Critical and Subcritical Fast Reactor Transmutation Systems

Fuel cycle analyses are performed to evaluate the impacts of further transmutation of spent nuclear fuel on high-level and low-level waste mass flows into repositories, on the composition and toxicity of the high-level waste, on the capacity of high-level waste repositories, and on the proliferation resistance of the high-level waste. Storage intact of light water reactor (LWR) spent nuclear fuel, a single recycle in a LWR of the plutonium as mixed-oxide fuel, and the repeated recycle of the transuranics in critical and subcritical fast reactors are compared with the focus on the waste management performance of these systems. Other considerations such as cost and technological challenges were beyond the scope of this study. The overall conclusion of the studies is that repeated recycling of the transuranics from spent nuclear fuel would significantly increase the capacity of high-level waste repositories per unit of nuclear energy produced, significantly increase the nuclear energy production per unit mass of uranium ore mined, significantly reduce the radiotoxicity of the waste streams per unit of nuclear energy produced, and significantly enhance the proliferation resistance of the material stored in high-level waste repositories.

Geochemical modelling of bentonite porewater in high-level waste repositories

The description of the geochemical properties of the bentonite backfill that serves as engineered barrier for nuclear repositories is a central issue for perfomance assessment since these play a large role in determining the fate of contaminants released from the waste. In this study the porewater chemistry of bentonite was assessed with a thermodynamic modelling approach that includes ion exchange, surface complexation and mineral equilibrium reactions. The focus was to identify the geochemical reactions controlling the major ion chemistry and acid–base properties and to explore parameter uncertainties specifically at high compaction degrees. First, the adequacy of the approach was tested with two distinct surface complexation models by describing recent experimental data performed at highly varying solid/liquid ratios and ionic strengths. The results indicate adequate prediction of the entire experimental data set. Second, the modelling was extended to repository conditions, taking as an example the current Swiss concept for high-level waste where the compacted bentonite backfill is surrounded by argillaceous rock. The main reactions controlling major ion chemistry were found to be calcite equilibrium and concurrent Na–Ca exchange reactions and de-protonation of functional surface groups. Third, a sensitivity analysis of the main model parameters was performed. The results thereof indicate a remarkable robustness of the model with regard to parameter uncertainties. The bentonite system is characterised by a large acid–base buffering capacity which leads to stable pH-conditions. The uncertainty in pH was found to be mainly induced by the pCO 2 of the surrounding host rock. The results of a simple diffusion–reaction model indicate only minor changes of porewater composition with time, which is primarily due to the geochemical similarities of the bentonite and the argillaceous host rock. Overall, the results show the usefulness of simple thermodynamic models to describe porewater chemistry of expandable clays although significant uncertainties with regard to the effects of swelling and physico-chemical properties of the interstitial water remain.