Nuclear reprocessing plants in Japan produce radioactive iodine-bearing materials such as spent silver adsorbents. Japanese disposal plans classify radioactive waste containing a given quantity of iodine-129 as Transuranic Waste Group 1 for spent silver adsorbent or as Group 3 for bitumen-solidified waste, and stipulate that such waste must be disposed of by burial deep underground. Given the long half-life of iodine-129 of 15.7 million years, it is difficult to prevent release of iodine-129 from the waste into the surrounding environment in the long term. Moreover, because ionic iodine is soluble and not readily adsorbed, its migration is not significantly retarded by engineered or natural barriers. The release of iodine-129 from nuclear waste therefore must be restricted to permit reliable safety assessment; this technique is called “controlled release”. It is desirable that the release period for iodine be longer than 100,000 years. To this end, several techniques for immobilization of iodine have been developed; three leading techniques are the use of synthetic rock (alumina matrix solidification), BPI (BiPbO2I) glass, and high-performance cement. Iodine is fixed as AgI in the grain boundary of corundum or quartz through hot isostatic pressing in synthetic rock, as BPI in boron/lead-based glass, or as cement minerals such as ettringite in high-performance alumina cement. These techniques are assessed by three models: the corrosion model, the leaching model, and the solubility-equilibrium model. This paper describes the current status of these three techniques.

Since 1991, in France, studies on the conditioning of iodine were carried out to assess the potential of several specific inorganic host matrices. The apatite family has been mainly studied because of its good chemical durability and its ability to confine iodine over geological time scales. A lead-bearing apatite, Pb10(VO4)4.8(PO4)1.2I2, and a calcium-bearing apatite, Ca10(PO4)6(OH)2-x(IO3)x, were selected on the basis of their incorporation rate (between 7 and 10 wt.%) and a satisfactory resistance to leaching (V0(50 °C, pure water) ∼ 10-2 g.m-2.d-1; Vr(50 °C, pure water) < 10-4 g.m-2.d-1). However, with such materials, the removal of open porosity requires non conventional sintering techniques like spark plasma sintering to decrease the surface exposed to water. This is why, in parallel, other matrices, like silver phosphate glasses, have also been investigated. To improve the chemical durability and thermal properties of these glasses, cross-linking reagents were added to their formulation.

Alumina matrix solidification is a hot isostatic pressing (HIP) technique used to immobilize radioactive iodine (129I) in the form of silver iodide. In the present study, an alumina matrix solidification sample with a porosity of 12.9% was obtained by performing HIP at 175 MPa and 1200°C for 3 hours on a simulated spent silver-sorbent saturated with stable iodine. Material Characterization Centre-1 (MCC-1) leaching tests for the simulated waste form were performed using hydrosulfide (HS-) as a reductant at concentrations ranging from 3 × 10-7 M to 3 × 10-3 M and at pH values ranging from 8.0 to 12.5. Leached iodine concentrations were below the detection limit for ICP-MS measurements at HS- concentrations of 3 × 10-7 M and 3 × 10-5 M. This result was due to the stability of AgI. At an HS- concentration of 3 × 10-3 M, iodine leaching rapidly increased within 10 days. The maximum iodine concentration in the solution was 4.33 × 10-3 M, which corresponds to 85% dissolution of the initial iodine. This value was measured after 552 days under an HS- concentration of 3 × 10-3 M at pH 11. An analysis of specimen cross-sections suggested the following reaction: 2AgI + HS- = Ag2S + 2I- + H+. The pH affected matrix aluminum dissolution but did not significantly affect the iodine leaching behavior. Furthermore, the normalized mass loss of iodine was larger than that of aluminum by a factor greater than 104, which is due to the large porosity and the dissolution of interior AgI of the solid.

The migration retardation of anionic radionuclides, notably I-129, in radioactive waste repositories is one of the most critical factors for improving the performance of engineered barriers. To gain more fundamental knowledge required to make such improvements, this study examined the sorption behavior of iodide ions on calcium silicate hydrate (CSH) and hydrotalcite (HT), which act as anion exchangers. CSH was synthesized using CaO and fumed silica, with Ca/Si molar ratios ranging from 0.4 to 1.6. The weight ratio of CSH to HT was 1.0. These solid samples were immersed for 14 days in a 30 mL sample of pure water or 0.6 M NaCl solution, each of which contained 0.5 mM iodide ions with a given liquid/solid weight ratio (10, 15, or 20). Raman spectroscopy studies indicated that the structures of CSH and HT were maintained during the hydration of the solid phase and the sorption of iodide ions. The distribution coefficients for the sorption of iodide ions on CSH and HT ranged from 6 to 13 L/kg for pure water and from 1 to 2 L/kg for NaCl solution. These retardation effects for iodide ions would contribute toward improving the performance of the repository system as most conventional safety assessments assume that iodide ions hardly sorb on engineered barriers such as cementitious materials.

For safety assessment analyses of the disposal of spent nuclear fuel (SNF) in deep geological repositories it is indispensable to evaluate the contribution of fission products to the instant release fraction (IRF). During the last three years the EURATOM FP7 Collaborative Project, “Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)” was carried out to get a better understanding of the IRF.

Within CP FIRST-Nuclides, a leaching experiment with a cladded SNF pellet was performed in bicarbonate water (19 mM NaCl + 1 mM NaHCO3) under Ar /H2 atmosphere over 333 days. The cladded SNF pellet was obtained from a fuel rod segment which was irradiated in the Gösgen pressurized water reactor; the average burn-up of the segment was 50.4 MWd/kgUO2. In the multi-sampling experiment, gaseous and liquid samples were taken periodically. The moles of the fission gases Kr and Xe released in the gas phase and those of 129I and 137Cs released in solution were measured. Cumulative release fractions of (1.6 ± 0.2)·10-1 fission gases, (1.6 ± 0.1)·10-1129I and (3.9 ± 0.2)·10-2 137Cs, respectively, were achieved after 333 days of leaching. Accordingly the release ratio of fission gases to 129I was 1:1 and the release ratio of fission gases to 137Cs was 4:1, respectively.

A cementitious waste form known as Cast Stone is the baseline waste form for solidification of aqueous secondary wastes, including Hanford Tank Waste Treatment and Immobilization Plant (WTP) secondary liquid effluents to be treated and solidified at the Hanford Site Effluent Treatment Facility. Cast Stone is also being evaluated as a possible supplemental immobilization technology to provide the necessary low activity waste (LAW) treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. Two radionuclides of particular concern in these waste streams are technetium-99 (99Tc) and iodine-129 (129I). These radioactive tank waste components are predicted to contribute the most risk to groundwater – the most probable pathway for future environmental impacts associated with the cleanup of the Hanford site. A recent environmental assessment of Cast Stone performance, which assumes a diffusion controlled release of contaminants from the waste form, calculates groundwater in excess of the allowable maximum permissible concentrations for both contaminants. There is, therefore, a need and an opportunity to improve the retention of both 99Tc and 129I in Cast Stone. One method to improve the performance of Cast Stone is through the addition of “getters” that selectively sequester Tc and I, therefore reducing their diffusion out of Cast Stone. In this paper, we present results of Tc and I removal from solution with various getters. Batch sorption experiments were conducted with deionized water (DIW) and a highly caustic LAW simulant with a 7.8 M average Na concentration. In general, the data show that the selected getters are effective in DIW but their performance is compromised when experiments are performed with the 7.8 M Na Ave LAW simulant. The diminished performance in the LAW simulant may be due to competition with Cr present in the 7.8 M Na Ave LAW simulant and to a pH effect that may create a negatively charged surface that can repel negatively charged species.

Calculations based on Poisson-Boltzmann theory are used to investigate the equilibrium properties of an electrolyte containing TcO4− and SO42− ions near the surface of amorphous silica. The calculations show that the concentration of TcO4− is greater than SO42− at distances less than 1 nm from the surface due to the negative charge density caused by deprotonation of the amorphous silica silanol groups. At lower pH, the surface becomes protonated and the magnitude of this effect is reduced. These results have implications for the potential use of oxyanion-SAMMS for the environmental remediation of water contaminated with 99Tc.

Sodalite (Na8[AlSiO4]6Cl2), a naturally occurring Cl-containing mineral, has long been regarded as a potential immobilization matrix for the chloride salt wastes arising from pyrochemical reprocessing operations, as it allows for the conditioning of the waste salt as a whole without the need for any pre-treatment. Here the consolidation and densification of Sm-doped sodalite (as an analogue for AnCl3) has been investigated with the aim of producing fully dense (i.e. > 95 % t.d.) ceramic monoliths via conventional cold-press-and-sinter techniques at temperatures of < 1000 °C. Microstructural analysis of pressed and sintered sodalite powders under these conditions is shown to produce poorly sintered, porous, inhomogeneous pellets. However, by the addition of a sodium aluminophosphate glass sintering aid, fully dense Sm-sodalite ceramic monoliths can successfully be produced by sintering at temperatures as low as 800 °C.

Alkaline earth aluminosilicate glasses (AeAS) with different MoO3 additions have been produced and assessed. MoO3 solubility increases with the equimolar substitution of smaller to larger alkaline earths and reaches 5.34 mol% in magnesium aluminosilicate glass (MAS). All visibly homogeneous glasses are X-ray amorphous, while the partially crystallised glasses exhibit some small X-ray diffraction peaks which are probably due to corresponding molybdates. The addition of MoO3 decreases glass transition and crystallisation temperatures and creates two broad Raman bands which are assigned to vibrations of MoO42‒ tetrahedra. The intensities of these bands increase along with MoO3 incorporation until the maximum solubility is reached. Electron microscopy shows that these separated particles are spherical, with sub-micron diameters and are randomly dispersed within glass. The separated phases are formed through liquid-liquid separation and thereafter crystallisation. Overall AeAS glasses look quite promising for molybdate immobilisation with MAS glasses being particularly attractive.

Two Mo-bearing glasses considered as candidate forms for high level waste (HLW) a uranium-graphite reactor spent nuclear fuel (SNF) reprocessing were characterized. Incorporation of Mo in sodium aluminophosphate (SAP) glass increases its tendency to devitrification with segregation of orthophosphate phases. Valence state and local environment of Mo in the materials containing ∼2 wt.% MoO3 were determined by X-ray absorption fine structure (XAFS) spectroscopy. In the quenched samples composed of major vitreous and minor AlPO4 nearly all Mo is located in the vitreous phase as [Mo6+О6] units whereas in the annealed samples Mo is partitioned among vitreous and one or two orthophosphate crystalline phases in favor of the vitreous phase. Mo predominantly exists in a hexavalent state in distorted octahedral environment. Four oxygen ions are positioned at a distance of ∼1.71-1.73 Å and two - at a distance of 2.02-2.04 Å. Minor Mo(V) is also present as indicated by a response in EPR spectra with g ≈ 1.911-1.915.

Copper valence and environment in two sodium aluminophosphate glasses suggested for immobilization of HLW from reprocessing of spent fuel of uranium-graphite channel reactor (Russian AMB) were studied by XRD, SEM/EDX, XAFS and EPR. Target glass formulations contained ∼2.4-2.5 mol.% CuO. The quenched samples were predominantly amorphous. The annealed MgO free sample had higher degree of crystallinity than the annealed MgO-bearing sample but both them contained orthophosphate phases. Cu in the materials was partitioned in favor of the vitreous phase. In all the samples copper is present as major Cu2+ and minor Cu+ ions. Cu2+ ions form planar square complexes (CN=4) with a Cu2+-O distance of 1.93-1.95 Å. Two more ions are positioned at a distance of 2.76-2.86 Å from Cu2+ ions. So the Cu2+ environment looks like a strongly elongated octahedron as it also follows from the absence of the pre-edge peak due to 1s→3d transition in Cu K edge XANES spectra of the materials. Cu+ ions form two collinear bonds at Cu+-O distances of 1.80-1.85 Å. Thus average Cu coordination number (CN) in the first shell was found to be 2.7-3.0.

The understanding of the crystallization of aluminosilicate phases in nuclear waste glasses is a major challenge for nuclear waste vitrification. Robust studies on the compositional dependence of nepheline formation have focused on large compositional spaces with hundreds of glass compositions. However, there are clear benefits to obtaining complete descriptions of the conditions under which crystallization occurs for specific glasses, adding to the understanding of nucleation and growth kinetics and interfacial conditions. The focus of this work was the investigation of the microstructure and composition of one simulant high-level nuclear waste glass crystallized under isothermal and continuous cooling schedules. It was observed that conditions of low undercooling, nepheline was the most abundant aluminosilicate phase. Further undercooling led to the formation of additional phases such as calcium phosphate. Nepheline composition was independent of thermal history.

In this study wet chemical methods combined with UV-Vis spectroscopy were performed to quantify Fe(II)/Fe(III) ratios and total iron content of quenched alkali alumino-boro-silicate (simulated nuclear waste) glasses, applying a colorimetric method. We report lessons learned from experimental challenges encountered associated with the colorimetric method, where 1,10 phenanthroline method is complexed with dissolved glass powder and the resulting solution measured for absorbance at 520 nm to determine Fe(II). To obtain total iron, the solution was then equilibrated with a mild reducing agent to chance all Fe to Fe(II), and the absorbance measured again at 520 nm. These absorbance values allowed for calculation of the Fe(II)/Fe(III) ratio, and the total iron content in the glasses. Total Fe measured is somewhat higher than as-batched target values for waste glasses, but very accurate for reference BCR-2G glass. All quenched alumino-boro-silicate glasses analyzed showed a Fe(II)/Fe(III) ratio between 0.06 (± 0.01) and 0.04 (± 0.01). These values are consistent with those obtained for similar glass compositions melted under analogous conditions, indicating a composition of ca. 94-96% Fe(III).

The determination of the long-term stability and corrosion of vitrified nuclear waste is an important aspect of research for the U.S. Department of Energy (DOE). It is necessary to understand the rate and mechanisms of Nuclear Waste Glass (NWG) corrosion to determine whether or not the glassy matrix will be able to retain radionuclides for the required repository performance time period. Glass corrosion and the rate of glass corrosions is determined by both chemical and microscopy. Electron Microprobe Analysis (EPMA) is a common and powerful method utilized in the examination of the chemographic difference between corroded and uncorroded NGWs. In this work, two forms of quantitative and semi-quantitative EPMA methods are defined by optimizing the instruments counting statistics against a standard glass and NIST minerals that have compositions similar to the glasses under examination. Data collected on both the planar and cross-sectioned surfaces of an unaltered simulated NWG by Standard based Wavelength Dispersive Spectroscopy (WDS) was found to be comparable to the theoretical composition of the glass. Conventional standardless Energy Dispersive Spectroscopy (EDS) data collected on the same surfaces was not comparable. However, standard-based EDS analysis is shown to be able to discriminate between unaltered and corroded glass surfaces.

To efficiently vitrify Hanford waste, the melting process (i.e., melter feed turning into waste glass) must be modeled and optimized. The rate of heat transfer to the melter feed in a waste glass melter, and thus the rate of melting, is strongly affected by the melter feed porosity, especially in the final stages where the glass-forming melt produces foam that insulates the feed from the molten glass. The volume expansion test allows the determination of the melter feed porosity as a function of temperature. This test measures the profile area of the feed pellet as it turns into glass. This contribution presents the calculation of the void fraction (porosity) of the melter feed as a function of temperature, heating rate, and material parameters. The process of finding the void fraction is described as well as results from the application of this process.

Calcium chlorosilicate (Ca3(SiO4)Cl2) is seen as a potential host phase for the immobilization of Cl-rich wastes arising from pyrochemical reprocessing, a waste stream often containing a mix of both di- and trivalent cations. Substitution of trivalent cations into the lattice requires some form of charge compensation to ensure the lattice remains charge neutral overall. Whilst previous work has only examined this through the formation of Ca vacancies, this study investigates the feasibility of charge-balancing via the substitution of a monovalent cation onto the Ca sites of the lattice. To that end, a series of static lattice calculations were performed to determine the site selectivity of monovalent cations of differing size when substituted onto the Ca sites of the calcium chlorosilicate lattice and the solution energies for the overall substitution processes compared with those for charge compensation via vacancy formation. In all cases the monovalent charge-balancing species shows a clear preference for substitution onto the Ca1 site in the calcium chlorosilicate lattice. The solution energy of the substitution process increases with the increasing ionic radii of both the mono- and trivalent species as the steric stresses associated with substitution of larger cations than the Ca2+ host increase. As such, only charge-balancing using Li+, Na+ or K+ is more favourable than via formation of a Ca vacancy.

Apatites are often seen as good potential candidates for the immobilization of halide-rich wastes. In particular, phosphate apatites have received much attention in recent years, however, their synthesis often produces complicated multi-phase systems, with a number of secondary phases forming [1.2]. Calcium vanadinite (Ca5(VO4)3Cl) demonstrates a much simpler phase system, with only a single Ca2V2O7 secondary phase which can easily be retarded by the addition of excess CaCl2. However, when doping with SmCl3 (as an inactive analogue for AnCl3) the Sm forms a wakefieldite (SmVO4) phase rather than being immobilized within the vanadinite, a result of having to form an energetically unfavourable Ca vacancy in order for the lattice to remain neutral overall. It has been postulated that charge-balancing the lattice via co-substitution of a monovalent cation will be less disfavoured and therefore help stabilise formation of a (Ca5-2xSmxAx)(VO4)3Cl solid solution (A = monovalent cation). This has been investigated using a combined modelling and experimental approach. Static lattice calculations performed using Li+, Na+ and K+ as charge-balancing species have shown the energy cost to be less than half that of charge-balancing via formation of a Ca vacancy. As a result, solid state synthesis of (Ca5-2xSmxLix)(VO4)3Cl, (Ca5−2xSmxNax)(VO4)3Cl and (Ca5-2xSmxKx)(VO4)3Cl solid solutions have been trialled, and analysis of the resulting products has shown a significant reduction in both the SmVO4 and Ca2V2O7 secondary phases across all dopant levels.

According to the Planning Act of 28th June 2006, Andra is in charge of ensuring the sustainable management of all radioactive waste generated in France, especially the high-level and long-lived vitrified waste produced from spent fuel recycling.

Since 2006, all the studies and research related to the components of HLW cells have been incorporated into a broader R&D program which aims at characterizing and modeling (i) the glass matrix dissolution, (ii) the corrosion of the overpack and the lining, and (iii) the claystone evolution in the near field, considering all the interactions between these surrounding materials. This program, coordinated by Andra, has involved up to eighteen laboratories.

After closure of disposal cells and overpack failure, glass alteration is expected to begin in partially saturated conditions due to hydrogen production resulting from carbon steel corrosion in anoxic conditions. Therefore, the glass should at least partially be hydrated by water vapor during thousands of years until complete saturation. A part of the studies aimed to determine the glass behavior in such conditions, the influence of the main parameters (temperature, relative humidity) and consequences of vapor hydration on subsequent radionuclides release by water leaching.

In addition, the major part of the work focused on the influence of the environment on glass alteration. The effect of clay pore water on glass alteration rates (initial rate, rate drop and residual rate) was determined and particularly that of pH and magnesium. The nature of steel corrosion products and their interactions with glass alteration were also investigated. All these studies relied on experiments in surface laboratories, in Andra’s underground laboratory, together with natural or archeological analogs and modeling studies.

A complete understanding of radioactive waste glass interactions with near-field materials is essential for appropriate nuclear waste repository performance assessment. In many geologic repository designs, Fe is present both in the natural environment and in the containers that will hold the waste glasses. In this paper we discuss investigations of the alteration of International Simple Glass (ISG) in the presence of Fe0 foil and hematite (Fe2O3). Based on solid analysis, ISG alteration is more pronounced in the presence of Fe0 than with hematite. Additionally, typical glass corrosion is observed for distances of 5 mm between Fe materials and ISG, but incorporation of Fe in the alteration layer is only observed for systems exhibiting full contact between Fe0 material and ISG. Solution analysis results indicate that diatomaceous earth minimizes corrosion to a larger extent than fumed silica does when present with iron and ISG.

Series of 7-day Product Consistency Tests (PCTs) were conducted with ARM-1 glass using the -100+200 mesh size fraction and several sub-fractions to measure the sensitivity of the test response to the distribution of particle sizes. Separate samples were prepared for testing by dry sieving and wet sieving, and the particle size distributions and PCT responses were measured for each fraction. Triplicate tests were conducted at 90 °C using a water/glass mass ratio of 10.0 with each size fraction. Test results are evaluated regarding the sensitivity of the test response to the particle size distributions and, conversely, the uncertainty due to calculating the surface areas (and dissolution rates) by modeling the particles as spheres. These analyses show the solution feedback effects of dissolved glass constituents (i.e., the reaction affinity) counteract the effects of the glass surface areas provided by different particle size distributions on the test response. The opposing effects of the surface area on the amount of glass dissolved and on the glass dissolution rate moderate the sensitivity of the PCT response to the particle size distribution.