Recent glass dissolution experiments, conducted at 90°C in the presence of potential backfill materials, indicate remarkably faster glass corrosion in the presence of clay, compared to tests where the glass is leached either alone or with alternative backfill materials. This effect correlates with the clay content in the backfill, and may be attributed to the removal of silica from solution. Sorption, or dissolution with reprecipitation of a silica-rich clay, have been proposed as possible mechanisms for the silica consumption.

The results of some experiments have been tested against a glass dissolution model, in which a widely used kinetic equation for glass corrosion is coupled with diffusive silica transport through a single porosity, linearly sorbing medium, which represents the backfilling. Because the glass corrosion rates imposed by the kinetic equation are inversely proportional to the silicic acid concentration of the leachant contacting the glass, the model predicts enhanced glass dissolution if silica is sorbed by the porous medium.

The experimental data proved to be consistent with the predicted enhancement of the glass dissolution. Moreover, the model-estimated distribution coefficients for silica sorption (Kd) fall within the range of values extracted from available literature data, thus supporting the hypothesis that the observed high corrosion rates are due to sorption of silica on the clay mineral surfaces.

Leach tests with R7T7 nuclear waste glass in distilled water were conducted at 50 and 90°C under static and slow flowing conditions, with an SA/V ratio of 50 m−1. A computer model for glass dissolution (LIXIVER) Was used to interpret the experimental data. This mechanistic model is based on a combination of the first-order law governing surface reactions, the silicon mass transport equation for the interstitial solution in the alteration film, and an empirical law for partial silicon retention in the alteration layer. The LIXIVER model satisfactorily accounts for most of the experimental results. The importance of the silicon diffusion mechanism in the diffusion layer is stressed. Values are indicated for the apparent silicon diffusion coefficient, Dsi, which ranged from 10−16 to 10−14 m2·s−1 at 50°C, and from 10−15 to 10−13 m2·s−1 at 90°C.

Experiments were conducted in initially pure water at 90°C and 100°C with glass specimens enriched in MgO and in alkali metals relative to the standard light water R7T7 reference glass. Three experimental protocols were implemented: Soxhlet testing with leachant renewal, static leaching according to a modified MCC-1 method with an SA/V ratio of 50 m−1, and static leaching with glass powder for an SA/V ratio of 8000 m−1 The results clearly show that the short and, especially, the long term dissolution rates depend on the initial glass composition. Higher MgO, Na2O, Li2O and B2O3 concentrations reduce the glass resistance to aqueous corrosion, as already indicated by the calculated hydration energy values. These experiments illustrate the importance of the glass composition in insuring long-term material integrity.

Several Savannah River Plant (SRL) glass compositions were reacted in steam at temperatures of 150 to 200°C. Half of the tests utilized actinide-doped monoliths and were exposed to an external ionizing gamma source, while the remainder were doped only with U and reacted without gamma exposure. All glass samples readily reacted to form secondary mineral phases within the first week of testing. An in situ layer of smectite initially developed on nonirradiated SRL 202 glass test samples. After 21 days, a thin layer of illite was precipitated from solution onto the smectite layer. A number of alteration products including zeolite, Casilicate, and alkali or alkaline earth uranyl silicate phases were also distributed over most sample surfaces. In the irradiated SRL 202 glass tests, up to three layers enveloped rounded, and sometimes fractured, glass cores. After 35 to 56 days these remnant cores were replaced by a mottled or banded Fe- and Si-rich material. The formation of some secondary mineral phases also has been accelerated in the irradiated tests, and in some instances, the irradiated environment may have led to the precipitation of a different suite of minerals. The alteration layer(s) developed at rates of 2.3 and 32 μm/day for the nonirradiated and irradiated SRL 202 glasses, respectively, indicating that layer development is accelerated by a factor of ∼10 to 15X due to radiation exposure under the test conditions.

A simulated nuclear waste glass was self-irradiated by doping with short-lived actinides of 238Pu and 244Cm. Changes in the hardness, the Young's modulus and the fracture toughness, as a function of irradiation dose, were measured by use of indentation techniques. The irradiated glass was annealed at temperatures from 573K to 723K for periods of up to 48hours, and the recovery of these changes were measured as a function of annealing temperature and time.

It was observed that the hardness and the Young's modulus decreased, while the fracture toughness increased exponentially with the cumulative dose. The maximum values of the relative changes in the hardness, the Young's modulus and the fracture toughness were about −25%, −30% and +45%, respectively. The results of the annealing show that the hardness and the Young's modulus were almost recovered to the original values at temperatures above 673K within 10 hours, while the recovery of the fracture toughness was minimal in this region of temperature and time. The changes in the hardness and the Young's modulus can be well explained by the model, in which the changes is proportional to the volume fraction of damaged zones, F, and the recovery of F is first order. On the other hand, the changes in the fracture toughness cannot be explained by the model, which suggests that the mechanism of the change in the fracture toughness is different from that in the hardness and the Young's modulus.

An Analytical Electron Microscopy study of colloidal particles formed during reaction of waste glass has been performed. The effect of waste glass test parameters on colloid formation is examined. Characterization of phases present in the leachate of these tests has shown that layers spalled from the glass and precipitated phases are both sources of colloids in the leachate. Elements, such as uranium, have been found to concentrate within colloidal particles in the leachate.

Static tests at SA/V (ratio of surface area of glass to solution volume) 20,000 m−1 on SRL 200 glass compositions show that, at long test periods, the simulated nuclear waste glass (nonradioactive) leaches faster than the corresponding radioactive glass by a factor of about 40, although comparative tests, done through 560 days, at lower SA/V, 2000 m−1, indicate little difference in the leach behavior of the two types of glasses. The similarity in leach behavior between radioactive and simulated glasses at SAN of 2000 m−1 or lower is also observed for SRL 165/42 and 131/11 compositions. The accelerated glass reaction with the simulated glass 200S is associated with the formation of crystalline phases such as clinoptilolite (or potassium feldspar), and a pH excursion. The radiation field generated by the fully radioactive glass reduces the solution pH. This lower pH, in turn, may retard the onset of increased reaction rate. The radiation field generated by the radioactive glasses does not directly affect the stability of the glass surface alteration layer under those conditions where the radioactive and simulated glasses react at the same rate. These results suggest that the fully radioactive nuclear waste glass 200R may maintain a much lower leach rate than the simulated 200S, if the lower pH in the 200R leachate can be sustained. Meaningful comparison tests between radioactive and simulated nuclear waste glasses should include long-term and high SA/V tests.

The interaction of borosilicate waste glasses with water has been studied extensively, and reasonably good models are available that describe the reaction kinetics and solution chemical effects. Unfortunately, these models have not been utilized in performance assessment analyses, except in estimating radionuclide solubilities at the waste form surface. A geochemical model has been incorporated in the AREST code to examine the coupled processes of glass dissolution and transport within the engineered barrier system. Our calculations show that the typical performance assessment analyses using fixed solubilities or constant reaction rate at the waste form surface are not always conservative or realistic predictions of radionuclide release. Varying the transport properties of the waste package materials is shown to produce counterintuitive effects on the release rates of some radionuclides. The use of noncoupled performance assessment models could lead a repository designer to an erroneous conclusion about the advantages of one waste package design or host rock setting over another.

With the imminent startup, in the United States, of facilities for vitrification of high-level nuclear waste, a document has been prepared that compiles the scientific basis for understanding the alteration of the waste glass products under the range of service conditions to which they may be exposed during storage, transportation, and eventual geologic disposal. A summary of selected parts of the content of this document is provided.

Waste glass alterations in a geologic repository may include corrosion of the glass network due to groundwater and/or water vapor contact. Experimental testing results are described and interpreted in terms of the underlying chemical reactions and physical processes involved. The status of mechanistic modeling, which can be used for long-term predictions, is described and the remaining uncertainties associated with long-term simulations are summarized.

Taking into account the neutron activation (Bq/kWht) and the environmental impact in terms of Ingestion Toxic Potential IT (m3 H2O/kWht) of radioactive wastes coming from several low-activation structural material candidates for future nuclear fusion power reactors, the contribution of the material chemical composition elements and impurities to the overall radioactivity and IT has been assessed. The chemical impurities, also at very low concentrations, have been ascertained to give, in many cases, a considerable contribution to the radwastes environmental impact, as in the case of SiC/SiC fiber composites materials or low-activation martensitic steels.

The orign of impurities origin has been identified and efforts to use this information in the material production processes for nuclear fusion application are suggested.

Cement matrices, doped with 137Cs, were investigated to establish their physicochemical properties, mainly 137Cs leach-rate. Experiments were carried out in leachant media with different degrees of aggressiveness on the cement matrix, e.g., distilled water, Na2SO4 and NaCl solutions. A series of six cement samples, which were orthocylindrically-shaped, were treated in each of three leachant media. For each sample, the surface area exposed to leaching relative to the total surface area was increased from 1:6 to 6:6, at increments of 1/6 of the total surface area. Expected higher 137Cs leach-rates of the samples immersed in the more aggressive environment show very specific behavior and functional relations with respect to the exposed surface area. Radionuclide leach-rate coefficients did not exhibit linear behavior. Instead, leach-rate values were best fit by exponential equations to describe cement matrix behavior in the leachants. Different empirical equations were established to describe 137Cs leach-rates in the various leachants as a function of the relative surface area of the cement matrix exposed to the leachant.

Cementitious materials provide an ideal geochemical environment (e.g., high pH pore fluids and large surface areas for sorption) for immobilizing nuclear waste. The inclusion of reducing agents, such as blast furnace slag (BFS) can immobilize radionuclides by forming of solid sulfide phases. Thermodynamic calculations using the MINTEQ geochemical computer code indicate that elemental sulfur present in BFS reacts with the highly mobile pertechnetate anion (TcO4−) anion to form an insoluble technetium sulfide phase (Tc2S7(8)). Initially, the waste form very effectively immobilizes technetium. However, as oxygen diffuses into the waste form, an outer zone of oxidized concrete and a shrinking core of reduced intact concrete develops. Oxidation of sulfur in the outer zone results in increased technetium concentrations in the pore fluid because Tc2S7(8) oxidizes to the mobile TcO4− anion. The TcO4− anion can then diffuse from the waste form into the environment.

A mathematical model that accounts for diffusion of oxygen into concrete coupled with oxidation of sulfur and sulfide to sulfate has been developed. This model assumes the existence of an oxidized outer layer of concrete surrounding a shrinking core of reducing intact concrete. A sharp boundary between the two zones moves slowly inward resulting in oxidation of Tc2S7(8) and subsequent release of TcO4− via aqueous diffusion in the concrete pore fluids. The model indicates that this mechanism results in a linear dependance of release with the square root of time similar to pure diffusion. In addition, the release of technetium is related to the inverse of the square root of the concentration of BFS, indicating that performance will significantly increases with the addition of approximately 20 percent BFS to the cement mix.

This paper is based on the results of an in-progress research project on Alkali-Activated Cement System at MRL. The objective of this research is to establish the potential for large volume use of fly ash and slag as main components of the cement system. Alkali-activated Fly ash-slag Cement (AFC) was studied as a matrix for immobilization of nuclear waste. AFC is characterized by high early strength, high ultimate strength, low porosity, lower solubilities of the hydrates, and high resistance to chemical corrosion as well as to freezing and thawing. All these advanced properties are particularly favorable to the immobilization the nuclear wastes.

The Nuclear Regulatory Commission (NRC) regulations 10 CFR Part 61, “Licensing Requirements for Land Disposal of Radioactive Waste,” regulate the disposal of radioactive waste and provides, among other stipulations, that class B and C low-level radioactive waste (LLW) be stabilized. This is intended to ensure that solidified waste does not structurally degrade and cause subsidence in the disposal unit's cover system. It is reasoned that deterioration of the waste form could adversely effect the stability of the burial site and lead to the release of radionuclides to the environment. Because of its apparent structural integrity, cement has been widely used as a binder to solidify LLW. However, the resulting preparations called pozzolanic cements are susceptible to failure due to the actions of stress and environment.

This paper presents data from the literature that document the significance of biologically mediated chemical attack on concrete, in general. Concrete is susceptible to aggressive reaction with acids (both mineral and organic) of natural and anthropogenic origin. If persistent, such reactions ultimately lead to structural failure. Groups of microorganisms have been identified that are capable of metabolically converting organic and inorganic substrates into organic and mineral acids.

The literature supports the conclusions that acid-producing bacteria of one type or another could be prevalent in all soils, even at depths expected for burial of LLW. Given the appropriate conditions of micro-environment and suitable substrate for growth, these bacteria will create conditions conducive to concrete deterioration. Growth substrates used by acid-producing bacteria (i.e., ammonia compounds, other reduced nitrogen compounds, sulfur and reduced sulfur compounds, reduced iron compounds, as well as organic carbon) could be naturally present in the disposal environment or be provided by the contents of the waste form.

Sufficient evidence is presented which demonstrates the potential for microbially-influenced deterioration of cement-solidified LLW. These data are the basis for the NRC initiating the development of appropriate tests to determine the resistance of cement-solidified LLW to microbiologically induced degradation.

Cement is used as a coating matrix for nuclear waste or as an engineered barrier of waste repositories situated in geological formations.

The effect of mineral acids excreted by bacteria (Thiobacillus) or organic acids produced by fungi, on the biodegradation of cement is discussed. Organic acids are quantitatively and qualitatively determined during growth of fungi over a two-year period. Even with high pH conditions, pH of the cement ≈ 11, growth of microorganisms occurs.

Biodeterioration of cement is expressed in terms of bioleaching velocity of calcium and is observed by electron microscopy.

The setting of cement is adversely affected by soluble borates. To reduce interference, the extent to which borate can be insolubilized has been investigated. One specific mechanism of insolubilization is by inclusion into ettringite. Ettringite, Ca6Al2(SO4)3(OH)12·26H2O, is a normal and stable constituent of Portland cement. It has an open but non-zeolitic framework. Borate can substitute partially or fully for sulfate.

Formation conditions, solubility and stability of borate ettringites, Ca6Al2(BO4)2-4(OH,O)12·26H2O, are characterised using XRD, IR, DTA, and SEM.

The potential durability of borate ettringites in a repository environment have been assessed by exposing it to Na-sulfate and Na-carbonate attack at different concentrations. Ion exchange occurs; back substitution of borate by sulfate is incomplete; high carbonate concentrations can, however, decompose borate ettringite.

On heat treatment up to 85°C the crystalline morphology and the OH arrangement of the structure are altered, but the X-ray powder pattern, and hence its structural framework are largely unaffected.

It is concluded that ettringite has potential to reduce the solubility of borate.

Concrete is a major component in many low-level radioactive waste (LLW) disposal facilities. The use of concrete is widespread because of its physical and structural properties and because it provides geochemical control on metal and radionuclide releases. Organic compounds are often disposed with radionuclides in LLW disposal facilities. Interactions between radionuclides and chelating agents must be evaluated to estimate mobility of radionuclides in concrete vaults. This paper quantifies the effects of two common organic components [citric acid and ethylenediaminetetraacetic acid (EDTA)] on radionuclide mobility in concrete barriers by using equilibrium geochemical calculations.

Equilibrium speciation calculations indicate that some radionuclides are chelated in groundwater (pH 7) but are destabilized in the highly alkaline (pH 13) concrete pore fluids. Radionuclides complexed by EDTA and citrate are replaced by calcium in the concrete pore fluids. In addition, the citrate nuclide complex reacts to form uncomplexed citrate in concrete pore fluids. The chemical performance of concrete LLW disposal facilities should not be compromised by small amounts of chelating agents disposed with some radionuclides. However, EDTA may form significant nickel and cobalt complexes above the pH important in the long-term service life of concrete barriers.

At the British Nuclear Fuels (BNFL) Sellafield reprocessing plant in the United Kingdom, cement grouts based on ground granulated blast-furnace slag (BFS) and ordinary Portland cement (OPC) are used extensively for immobilising radioactive wastes. These grouts have excluded organic admixtures in order to reduce process complexity and uncertainties, regarding the performance of organic admixtures with BFS/OPC grouts, particularly under irradiation.

This study has investigated the effects of sulfonated melamine formaldehyde and naphthalene condensates on grout properties. The results show grout settlement and strengths increase on addition of additives, with the additives remaining largely in the pore solution. Under irradiation the additives breakdown liberating hydrogen and carbon dioxide. Strength and product dimensions are unaffected by irradiation.

This work is based upon the notion that a grid of electrodes will be incorporated in the slurry of radioactive waste and grout-forming solids. These electrodes will remain in the hardened grout, and will serve to monitor the ionic concentration of the liquids contained within the matrix, the structural integrity of the matrix, and the rates of out-diffusion of the electrolytes. Ultimately, we will investigate four techniques: (1) ionic conductivity, (2) electrochemical potential, (3) cyclic voltammetry, and (4) impedance measurements. Eventually, the results will be correlated with present quality verification methods, which include recovering samples and following the Toxicity Characteristic Leach Procedure (or other such accepted standard tests). This paper presents the preliminary results of a recently initiated effort.

The mining industry currently uses fly-ash from coal-fired power-plants to close access shafts in abandoned mines, in sedimentary formations. The technique consists of installing a plug of fly-ash, about fifty meters high, at the base of the shaft-lining. In this study, fly-ash is considered as a possible sealant for radioactive repositories in sedimentary formations, as well as in hard rock.

Specific studies have been conducted in order to assess the feasibility of this technique, and to improve the long-term performance of the fly-ash seals. The hydraulic and mechanical characteristics of fly-ash, taken from the plug of an abandoned shaft, as well as from dumps of various ages, were measured. Laboratory tests showed that it is possible to reduce the hydraulic conductivity of fly-ash by a factor of thousand, with the addition of ten percent bentonite. Moreover, it seems possible to block fine fissures in the surrounding damaged rock by injecting grouts made of cement and fly-ash having maximum diameters of twenty micrometers.