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In the context of the geological disposal of vitrified radioactive waste, cracks are expected at the full scale of the glass blocks. To simulate glass alteration in cracks, static dissolution experiments were conducted with confined SM539 glass powder, at 30 °C in a synthetic young cementitious water. Short and long-term glass dissolution rates were determined and compared with those measured in similar experiments with dispersed glass powder. A long-term rate ~ 10 times lower was found, which might be due to a limited access of the solution to the glass surface area, resulting in a much higher local SA/V and faster increase of local concentrations of leached glass constituents, with possibly a lower local pH. In both types of experiments, the altered glass powder was characterised by XRD and SEM-EDX. Some glass particles showed the presence of an amorphous multi-layer and of fibrous and spherical secondary phases. Graphical abstract

Cr-doped UO2-based model materials were prepared at SCK CEN, mimicking modern LWR fuels, to understand the influence of Cr doping on the spent fuel dissolution behaviour in geological repository conditions. Tests were carried out with four model materials: depleted UO2, Cr-doped depleted UO2, Pu-doped UO2 and Pu-Cr-doped UO2. Static dissolution experiments have been performed up to 4 months in autoclaves under 10 bar H2 pressure with a Pt/Pd catalyst in media at pH 13.5 and at pH 9. The Cr-doping appeared to reduce the U concentrations by a factor 6 at pH 13.5, but it had no or not much effect at pH 9. Graphic abstract

Autoclave leaching experiments are conducted on three well-characterised, irradiated, and cladded mixed oxide fuel-rod segments with burnups ranging from 29 GWd/tHM to 52 GWd/tHM to investigate the instant release fraction of fission gases and long-lived fission products and to assess the long-term fuel matrix corrosion. The segments are exposed to bicarbonate solutions as reference groundwater at neutral pH and a synthetic young cementitious water at pH 13.5 under reducing atmosphere (4 vol% H2 in Ar at 40 bar pressure), since 2018. The initial leaching results for the fission products caesium and iodine as representative elements of the instant release fraction were found to depend on the leachate composition as well as on the fuel burnup. Graphical abstract

Theslides werepresented at the 45th Scientific Basis for Nuclear Waste Management SymposiumonOctober 27, 2021. Abstract Disposal in a deep geological repository (DGR) based on a multibarrier concept is considered as the safest and most sustainable option for the management of spent nuclear fuels (SNF) in many countries. Demonstrating the long-term safety of a DGR for SNF over assessment time frames of up to one million years requires a profound understanding of the corrosion behaviour of SNF coming into contact with groundwater, when the waste canister is eventually breached. During the last decades, a number of studies addressed this topic, leading to a good phenomenological understanding of the long-term behaviour of SNF in a DGR. However, various processes contributing to the (radiolytic) matrix corrosion of SNF in the generally reducing repository environment are still not fully understood, and corrosion data on used mixed oxide (MOX) fuels are scarce to date. To investigate the effects of environmental conditions on the corrosion of SNF, the SF-ALE project (Spent Fuel Autoclave Leaching Experiments) was initiated as collaboration between the Belgian Nuclear Research Centre SCKCEN, Forschungszentrum Jülich GmbH, and the Belgian Agency for Radioactive Waste and Fissile Materials (NIRAS/ONDRAF). Within SF-ALE, the release of various radionuclides and fission gases from well-characterised, irradiated MOX fuel segments (burn-up between 29GWd/tHM and 52GWd/tHM) exposed to bicarbonate water as reference groundwater at neutral pH and a synthetic cementitious water at pH13.5 under reducing atmosphere (4vol% H2 in Ar at 40bars pressure) is investigated. The first project phase lasted for about two years and was completed in September 2020. The leaching solutions were renewed and a second leaching phase was started on the same MOX fuel segments, aiming for a total duration of about 3.5years. In this contribution, we report the results obtained during the first phase and describe the transition to phase two of the project.

Understanding the corrosion of nuclear waste glass is critical to predicting its safe disposal within a geological facility. The corrosion mechanisms and kinetics of the International Simple Glass, a simplified version of high-level nuclear waste glass, was shown to be significantly influenced by a high pH cement solution representative of disposal conditions. We provide the first microscopic characterisation of the porous, Zr-rich aluminoalkali-silica gel corrosion layer that was observed. Ca, Na and K from the cement solution were incorporated into the corrosion layer to charge compensate Si, Al and Zr species; the incorporation of Al was postulated to result in precipitation of an aluminosilicate-rich gel with large voids, facilitating rapid transport of species through the gel layer and significantly enhancing the corrosion rate. Precipitation of Al-containing zeolite and phyllosilicate phases was also observed, indicating that cementitious solutions may promote the detrimental ‘rate resumption’ stage of glass corrosion. The corrosion behaviour of nuclear-waste glass in solutions representative of cement-rich disposal facilities has been studied. Many countries have chosen to manage the waste created by nuclear-power generation by vitrifying and storing it in underground geological repositories, where cement will be used in general construction and, for example, to line vaults. Now, Claire Corkhill at the University of Sheffield, and collaborators in Belgium and the USA, have studied the mechanisms and kinetics of the dissolution of ‘International Simple Glass’ (ISG) in a cementitious solution, modelling the solution chemistry and characterising the surfaces of glass powders and monoliths using spectroscopic, diffraction and microscopic techniques. Their data suggest that, under these conditions, alkali/alkaline-earth aluminosilicate alteration layers formed in high-pH cement solutions significantly affect solute transport to and from the glass interface, thus enhancing the degradation of ISG.

Immobilization of high-level and intermediate-level nuclear wastes by vitrification in borosilicate glass is a well-established process. There is a consensus between the waste management agencies of many countries and many experts that vitrified nuclear waste should be disposed of in a deep geological waste repository and therefore its long-term behavior needs to be taken into account in safety assessments. In contact with water, borosilicate glass is metastable and dissolves. In static dissolution experiments, often a surface alteration layer (SAL) forms on the dissolving glass, and later sometimes secondary phases form. Based on boron or lithium release rates, commonly three stages of glass dissolution are defined as a function of the reaction progress: (I) initial dissolution, described by a congruent glass dissolution at the highest rate, (II) residual dissolution, characterized by a glass dissolution rate several orders of magnitude lower than the initial one, and (III) resumption of glass alteration with initial rates. Microscopically, the formation of a complex SAL has been identified as a prerequisite for the slower dissolution kinetics of stage II. Stage III is typically observed under specific conditions, i.e., high temperature and/or high pH driven by the uptake of Si and Al into secondary phases. Different glass dissolution models explaining the mechanisms of the SAL formation and rate-limiting steps have been proposed and are still under debate. In this article different aspects of glass dissolution from recent studies in the literature and our own work are discussed with a focus on the microscopic aspects of SAL formation, secondary phase formation and the resumption of glass dissolution. Most of the experiments in the literature were performed under near-neutral pH conditions and at 90 ∘C , following standard procedures, to understand the fundamental mechanisms of glass dissolution. The example of interaction of glass and cementitious materials as discussed here is relevant for safety assessments because most international concepts include cement e.g., as lining, for plugs, or as part of the general construction of the repository. The aim of the investigations presented in this paper was to study the combined effect of hyperalkaline conditions and very high surface area/volume ratios ( SA / V = 264 000 m - 1 ) on the dissolution of international simplified glass (ISG) and the formation of secondary phases at 70 ∘C in a synthetic young cement water containing Ca (YCWCa). The new results show that the SA / V ratio is a key parameter for the dissolution rate and for the formation of the altered glass surface and secondary phases. A comparison with similar studies in the literature shows that especially on the microscopic and nanoscale, different SA / V ratios lead to different features on the dissolving glass surface, even though the SA-normalized element release rates appear similar. Zeolite and Ca-silicate-hydrate phases (CSH) were identified and play a key role for the evolution of the solution chemistry. A kinetic dissolution model coupled with precipitation of secondary phases can be applied to relate the amount of dissolved glass to the evolution of the solution's pH.

Static dissolution experiments were carried out with the reference International Simple Glass under hyperalkaline pH at 70 °C and very high SA/V ratio. Three aspects of glass dissolution behavior were investigated, (1) the rate drop regime and the residual rate (stage II), (2) the formation of secondary phases including thermodynamic aspects, and (3) the microstructure of the interface of altered glass and secondary phases. A very low residual rate of 6 × 10−6 g/m2d was determined based on boron release, which was several orders of magnitude lower than the initial rate established between the start of the experiments and the first sampling on day 59. The presence of a porous layer with a thickness varying between 80 nm and 250 nm and a pore size between 10 nm and 50 nm was observed. CSH phases with a low Ca/Si ratio of 0.3–0.4 and zeolites were also visible at the surface of the altered glass grains, but no glass alteration resumption occurred, probably due to an important pH decrease already at day 59. Thermodynamic calculations assuming congruent glass dissolution and precipitation of the dissolved aqueous species confirmed the precipitation of CSH phases and zeolites.

Technical paper submitted for peer-review, accepted, and presented at the 2021 Topfuel conference (October 24-28, 2021). The conference was hosted as a hybrid event with on-site attendance (Santander, Spain) and remote attendance via a web application. ---Abstract--- In the context of interim or long-term storage of spent nuclear fuel, as well as in final disposal scenarios it is crucial to understand and accurately predict the materials��� behavior, specifically towards the release of radionuclides into the environment. Within the Spent Fuel Autoclave Leaching Experiments (SF-ALE) project at SCK CEN an improved experimental setup for performing leaching studies was developed, and six experiments (three on UOX fuel, three on MOX fuel) are being conducted in parallel over a timeframe of 36 months. To allow an accurate interpretation of the leaching data, depletion calculations were performed to establish the nuclide inventory of the MOX fuel samples at the time of the experiments. Preliminary results of the fission gas release measured during the leaching experiments are reported and discussed., {"references":["B.J. Lewis, W.T. Thompson, F.C. Iglesias, \"Fission Product Chemistry in Oxide Fuels\", in: R.J.M. Konings (Ed.) Comprehensive Nuclear Materials, Elsevier, Oxford, 2012, pp. 515-546.","D. Cubicciotti, J.E. Sanecki, Journal of Nuclear Materials, 78 (1978) 96.","H.J. Matzke, Radiation Effects, 53 (1980) 219.","R.G.J. Ball, W.G. Burns, J. Henshaw, M.A. Mignanelli, P.E. Potter, Journal of Nuclear Materials, 167 (1989) 191.","H. Kleykamp, Journal of Nuclear Materials, 171 (1990) 181.","T.M. Besmann, J.W. McMurray, S. Simunovic, Calphad - Computer Coupling of Phase Diagrams and Thermochemistry, 55, Part 1 (2016) 47.","E.H.P. Cordfunke, R.J.M. Konings, Journal of Nuclear Materials, 152 (1988) 301.","R.W. Grimes, C.R.A. Catlow, Philosophical transactions - Royal Society. Mathematical, Physical and engineering sciences, 335 (1991) 609.","H. Kleykamp, Journal of Nuclear Materials, 131 (1985) 221.","L. Johnson, I. Günther-Leopold, J. Kobler Waldis, H.P. Linder, J. Low, D. Cui, E. Ekeroth, K. Spahiu, L.Z. Evins, Journal of Nuclear Materials, 420 (2012) 54.","E. González-Robles, et al., Journal of Nuclear Materials, 479 (2016) 67.","D. Serrano-Purroy, F. Clarens, E. González-Robles, J.P. Glatz, D.H. Wegen, J. de Pablo, I. Casas, J. Giménez, A. Martínez-Esparza, Journal of Nuclear Materials, 427 (2012) 249.","D. Serrano-Purroy, I. Casas, E. González-Robles, J.P. Glatz, D.H. Wegen, F. Clarens, J. Giménez, J. de Pablo, A. Martínez-Esparza, Journal of Nuclear Materials, 434 (2013) 451.","K. Lemmens, et al., Journal of Nuclear Materials, 484 (2017) 307.","T. Mennecart, G. Leinders, C. Cachoir, G. Cornelis, G. Verpoucke, G. Modolo, D. Bosbach, K. Lemmens, M. Verwerft, \"First phase of the Spent Fuel Autoclave Leaching Experiments (SF-ALE) at SCK•CEN\", Global Top Fuel 2019, Seattle (USA) (2019).","T. Mennecart, K. Lemmens, C. Cachoir, L. Adriaensen, A. Dobney, \"Leaching tests for the experimental determination of IRF radionuclides from Belgian high-burnup spent nuclear fuel: Overview of pre-test fuel characterization experimental set-up and first results\", 2nd Annual Workshop Proceedings of the Collaborative Project \"Fast/Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel\", KIT Scientific report KIT- 7676, (2013).","B. Kienzler, V. Metz, L. Duro, A. Valls, \"State of the art report - Update 2013 of the Collaborative Project Fast/Instant Release of Safety Relevant Radionuclides from Spent Nuclear (7th EC FP CP FIRST-Nuclides)\", (2013).","L. Adriaensen, A. Dobney, P. Van Bree, \"Radiochemical burnup analysis of fuel sample FT1X57-D05-BU1\", SCK CEN Restricted contract report R-5640, (2014).","Wieselquist, W.A., THE SCALE 6.2 ORIGEN API FOR HIGH PERFORMANCE DEPLETION (2015)."]}

For the safety assessment of Spent Nuclear Fuel (SNF) in disposal conditions, it is crucial to determine the fast release of the potentially critical radionuclides (e.g. 129I, 137Cs, 99Tc, 93Mo, 90Sr…) upon contact with water. To underpin the quantification of this fast release, leaching experiments were performed at SCK CEN on a PWR UOX fuel from the Belgian Tihange 1 reactor, with a rod average burnup of 50.5 MWd.kg−1HM, an estimated fission gas release of 14% and an average linear power rate of 321 W.cm−1. Tests were carried out with durations up to one year, with clad and declad spent fuel samples, in a NaCl/NaHCO3 solution at room temperature for a period of one year, in a set-up designed to prevent the intrusion of airborne oxygen. In the test with the declad segment, the exposed surface area was much larger than in the test with the clad segment. As a result, in the first days of leaching, the release of all radionuclides was larger for the declad segment than for the clad segment, except for Tc. After one year, the cumulated release from the declad and clad segments are similar, except for Tc, as a result of air intrusion leading to a higher oxidation and concomitant dissolution resumption, more pronounced for the clad than for the declad segment. For the declad segment, the Tc and Mo concentrations decrease, suggesting reduction to less soluble species in spite of the air intrusion. The release of 129I is similar to the fission gas release, while the release of 137Cs is lower, but still accelerated compared to the UO2 matrix. The release of 99Tc, 95-98Mo (representative of 93Mo) and 90Sr is only slightly higher than the matrix dissolution rate. All determined released fractions are in line with the reported values of the literature.

The International Simple Glass (ISG) is considered as reference benchmark glass and is used in the frame of an international collaboration for the study of the dissolution mechanisms of high-level vitrified nuclear waste.In this work the forward dissolution rate of the ISG was determined in different alkaline solutions, as a simulation of the disposal conditions foreseen by the Belgian concept for geological disposal of vitrified waste. The determination of the forward dissolution rate was done by measuring the Si released from the glass in solution in tests performed at 30 °C in four different KOH solutions with pH varying from 9 to 14 and in artificial cementitious water at pH 13.5.The forward dissolution rates determined for the ISG in high pH solutions are in good agreement with the results obtained for a lower pH range.The rates obtained in this study, moreover, were compared with the rates measured in the same conditions for SON68 glass from a previous work. The values obtained for the two glasses are comparable in artificial cementitious water and in KOH at moderately alkaline pH. At higher pH, ISG glass shows a lower forward dissolution rate with respect to SON68 (0.20 g•m−2•d for ISG and 0.35 g•m−2 d for SON68 at pH 14).

In the framework of the Euratom FIRST-Nuclides program, SCK•CEN conducted leaching experiments on a PWR UOX fuel, taken from the Belgian Tihange 1 reactor, with an average rod burnup of 50 MWd/kgHM in order to determine the rapid release of some of the most critical radionuclides. Experiments were conducted with an intact and a declad spent fuel segment in contact with a reference carbonate solution. During one year, samplings were performed to determine the release of the most relevant radionuclides. The samples taken after six months of leaching were analysed by means of Thermal Ionisation Mass Spectrometry (TIMS). These results were compared to the total inventory measured with radiochemical burnup analyses, as well as with the calculated inventory profile along the pellet radius. The comparison gives information about the origin of the dissolved uranium.

Leaching experiments were performed with depleted UO2 powders in tetramethylammonium solutions (TMA-OH) at pH 13.5 and 12.5, and at different UO2 surface area to volume of solution (SA/V) ratio's to determine the solubility and the dissolution kinetics of UO2 at high pH in absence of cations dominating cementitious waters (Ca,Na, K). The solubility of UO2 increased from pH 12.5 to 13.5 and by increasing the SA/V ratio up to 100 m-1. However, no known U secondary-phases were predicted by geochemical calculations to control the measured U-concentrations. A two-step dissolution process was put forward: 1-a fast initial rate, surface controlled and hydroxo promoted and 2-a sorption process at low SA/V ratio or a continuous residual dissolution process at high SA/V ratio.

Understanding the dissolution of nuclear waste glass in hyper-alkaline conditions is critical to predicting its safe disposal in cement-dominated disposal conditions. In this work, solution data from the early-stage SON68 glass dissolution in a KOH solution (pH 13.5 and 30 °C) were interpreted with models based on silicon (Si) diffusion controlled stoichiometric dissolution of the glass. The solution data show a linear increase of boron (B) concentration with the square root of time and a decrease of the glass dissolution rate with an increase in the ratio of glass surface area to solution volume (SA/V). Three models from the literature, based on either aqueous pore diffusion or solid diffusion with/without resistance of the surface boundary layer, can fit the evolution of solution concentrations for the major elements of the glass equally well. However, the fitted Si diffusion coefficients differ by several orders of magnitude. Since all these fitted diffusion coefficients are lower than diffusion coefficients generally reported for aqueous pore diffusion, they are more indicative of solid diffusion processes. These results show that even under such hyper-alkaline conditions where dissolution and precipitation reactions are usually considered to prevail over solid diffusion, solid diffusion can still be an important dissolution mechanism. Those two dissolution mechanisms do not necessarily contradict each other. It is postulated that incorporation of potassium (K) from the solution onto the glass surface provides a diffusion barrier against the release of glass components into the solution. Therefore, the early-stage glass dissolution rate under hyper-alkaline conditions should be lower than the predicted glass network dissolution rates extrapolated based simply on pH effects.

We report the modelling and experimental studies of SON68 glass dissolution in hyperalkaline solution (pH = 13.7) at 30 and 70 °C for more than 300 days. In the experiments at 30 °C and the first 100 days at 70 °C, the concentrations of B, Mo, alkalis and even Si showed square root of time dependence indicative of a diffusion-controlled process. At 70 °C fast dissolution resumed after about 100 days due to crystallization of aluminosilicates, and glass dissolution became surface reaction-controlled. For the experiment at 30 °C, a simple model coupling dissolution of the glass and precipitation of secondary phases can also produce a good result for non-conservative elements, such as Ca and Al.

The instant release of fission products from high burn-up UO 2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45–63 GWd/t HM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride – bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H 2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

To assess the long-term behaviour of spent fuel in conditions of the Belgian Supercontainer design, dissolution tests were performed to determine the solubility and dissolution rate of depleted and Pu-doped UO2 in synthetic cement waters. The results confirm the behaviour generally observed with UO2 at neutral pH in absence of H2. Due to the radiolysis of water, the produced oxidizing species oxidise the U(IV) to U(VI), inducing an increase of the uranium concentration. Consequently, the dissolution rate increases from the depleted UO2 to the most active doped UO2. Imposing a H2 overpressure had no significant effect on the UO2 behaviour.

To assess the stability of spent fuel in the highly alkaline chemical environment of the Belgian Supercontainer design, static leach experiments were performed with depleted UO2 and 238Pu-doped UO2 at different SA/V ratios for 1.5 years in cement waters (11.7< pH < 13.5) at ambient temperature and under argon atmosphere. The influence of the calcium concentration on the uranium release was also investigated. While the ultrafiltered U(IV) concentration was 10-9-10-8 mol.L-1 and independent of the pH, the U(VI) release from the UO2 surface was enhanced by the OH- concentration, leading to soluble U concentrations up to 10-5 mol.L-1 at high SA/V and under the influence of radiolysis. Together with the high Ca concentration, this can lead to the formation of Ca-U(VI) colloids as precursor of Ca-U(VI) secondary phases, decreasing the soluble U concentration. The precipitation of Ca-U secondary phases was however not clearly evidenced by surface analyses.

The interaction between simulated reference waste glasses SON68 and SM539, and cement has been studied in suspensions of ordinary Portland cement and synthetic cement water with pH 13.5 at 30°C. The cement appears to trigger glass dissolution by consumption of glass matrix components. This leads to fast glass dissolution at a constant rate with formation of a porous gel layer on the glass. A glass dissolution model has been proposed considering that the silicon coming from the glass reacts with portlandite to form C-S-H phases. The transformation of C-S-H into C-A-S-H phases is a second parallel driving force especially for the Al-rich SM539 glass. After consumption of the portlandite, the glass alteration rate is expected to decrease.

At the appropriate times, silica diffusion in clay is possibly the rate determining process for the dissolution of vitrified waste disposed of in a clay layer. For testing this hypothesis, combined glass dissolution/silica diffusion experiment are performed. SON68 glass coupons doped with the radioactive tracer 32 Si are sandwiched between two cores of humid Boom Clay, heated to 30 °C. Due to glass dissolution, 32 Si is released and diffuses into the clay. At the end of an experiment, the mass loss of the glass coupon is measured and the clay core is sliced to determine the diffusion profile of the 32 Si released from the glass in the clay. Both mass loss and the 32 Si diffusion profile in the clay are described well by a model combining glass dissolution according to a linear rate law with silica diffusion in the clay. Fitting the experiments to this model leads to an apparent silica diffusion coefficient in the clay between 7 × 10 −13 m 2 /s and 1.2 × 10 −12 m 2 /s. Previously determined values from diffusion experiments at 25 °C are around 6 × 10 −13 m 2 /s (In-Diffusion experiments) and 2 × 10 −13 m 2 /s (percolation experiments). The maximal glass dissolution rate for glass next to clay is around 1.6 × 10 −7 g glass/m 2 s (i.e. 0.014 g glass/m 2 day). In undisturbed clay, the measured silica concentration is around 5 mg/L. Combining these values with the previously measured (In-Diffusion experiments) product of accessible porosity and retardation factor, leads in two ways to a silica glass saturation concentration in clay between 8 and 10 mg Si/L. Another candidate for the rate determining process of the dissolution of vitrified waste disposed in a clay layer is silica precipitation. Although silica precipitation due to glass dissolution has been shown experimentally at 90 °C, extending the model with silica precipitation does not lead to much better fits, nor could meaningful values of a possible precipitation rate be obtained.

An overview is given about the types of migration experiments in clay and their corresponding mathematical models aimed to determine: (1) the product ηR of the diffusion accessible porosity η and the retardation factor R, and (2) the apparent dispersion coefficient D. An example shows that not all set-ups allow to estimate values for these parameters. An important concern is consistency. Consistency requires a good correlation between the diffusion accessible porosity η (fitted from migration experiments) of tritiated water and the independently measured water content. Column migration experiments (pulse injection type) on compact clay cores taken at different depths in the Ypres Clay Formation, confirm this correlation and lead to the expected value of the grain density, showing that nearly the entire porosity of the clay is accessible for water diffusion. Another requirement for consistency is that different experimental set-ups lead to the same value for the migration parameters. The average values of the apparent diffusion coefficient of dissolved silica obtained with different types of tests in Boom Clay are, respectively, 6 × 10−13 m2/s (in-diffusion, 25 °C), 2 × 10−13 m2/s (percolation, 25 °C) and 9 × 10−13 m2/s (combined glass-dissolution/diffusion, 30 °C). The values of the product ηR are around 60–70 (in-diffusion) and between 100 and 150 (percolation). Using fit results from the combined glass-dissolution/diffusion experiments with other results, leads to ηR values ranging from 40 to 90. The values of the migration parameters obtained from different types of experiments and the corresponding models are consistent.

The solubility of uranium dioxide (UO2) was measured in real and synthetic Boom Clay waters with varying concentrations of humic acids and carbonate under reducing conditions at 20 C. Uranium concentrations in function of time suggest the reduction of U(VI) to U(IV) by the humic acids which is occurring faster in real clay water than in synthetic clay waters. Humic acids induce also a competition to complex U(VI) in carbonate-containing solution, but they are not able to control the uranium concentration at high bicarbonate concentration (0.02 mol dm � 3 ). Nevertheless they may play a role at low carbonate concentration. In our experimental conditions, the geochemical calculations indicate that two uranium secondary phases (U4O9 and UO2ðcÞ) are susceptible to control the uranium concentration in solution. These calculations are in good agreement with results of the X-ray photoelectron spectroscopy. At the end of tests, uranium concentrations reach steady-state values between 3 � 10 � 8 and 5 � 10 � 8 mol dm � 3 in the bicarbonate-rich solutions. Although these concentrations are considered as conservative, they are 10–100 times higher than in natural Boom Clay. The consequence is that spent fuel could slowly dissolve in the interstitial clay water undersaturated with respect to UO2/UO2þx of the fuel. 2003 Elsevier B.V. All rights reserved.

In the frame of the Safety and Feasibility Case 1 development, the Belgian agency for radioactive waste (ONDRAF/NIRAS) has to demonstrate that the geological disposal of spent nuclear fuel in the Boom Clay host rock following the Supercontainer design provides sufficient radiological safety in the long term. This safety relies on the containment of the radionuclides by the overpack, the limitation of radionuclide release from the waste matrix, the diffusion controlled transport of the radionuclides, and their low solubility and sorption on the Boom Clay host rock. This paper presents an evaluation of the main processes considered for the limitation of radionuclide release by the Spent Fuel in Supercontainer conditions, characterized by the presence of a concrete buffer and hyperalkaline pore water. We present a description of the main expected processes with their potential impact on the fuel alteration mechanisms, and potential reference fuel matrix alteration rates and effective fuel surface areas for use in safety and performance assessment models.

To assess the long-term behavior of spent fuel in alkaline conditions representative for the Belgian Supercontainer design, static and dynamic dissolution tests were performed with depleted and Pu-doped UO2 , simulating medium burn-up UOX fuels of different fuel ages. The experiments were performed under argon atmosphere at 25 – 30°C in cement waters in the pH range 11.7 – 13.5 and at different SA/V ratios. This paper presents the observed UO2 matrix dissolution rates based on the (238U or 233U) release, and proposes a selection of reference dissolution rates for performance assessment. We demonstrate that the dissolution rates at high pH are equivalent to the dissolution rates reported in the literature for neutral pH conditions. The α-activity threshold below which radiolytical fuel oxidation becomes negligible, seems to be close to the threshold reported for anoxic media at neutral pH.

The Belgian Nuclear Research Centre (SCK•CEN) has a long-standing expertise in research concerning the compatibility of waste forms with the final disposal environment. For high level waste, most attention goes to two waste forms that are relevant for Belgium, namely (1) vitrified waste from the reprocessing of spent fuel, and (2) spent fuel as such, referring to the direct disposal scenario. The expertise lies especially in the study of the chemical interactions between the waste forms and the disposal environment. This is done by laboratory experiments, supported by modeling. The experiments vary from traditional leach tests, to more specific tests for the determination of particular parameters, and highly realistic experiments. This results in a description of the phenomena that are expected upon disposal of the waste forms, and in quantitative data that allow a conservative long-term prediction of the in situ life time of the waste form. The predictions are validated by in situ experiments in the underground research laboratory HADES. The final objective of these studies, is to estimate the contribution of the waste form to the overall safety of the disposal system, as part of the Safety and Feasibility Case, planned by the national agency ONDRAF/NIRAS. The recent change of the Belgian disposal concept from an engineered barrier system based on the use of bentonite clay to a system based on a concrete buffer has caused a reorientation of the research programme. The expertise in the area of clay-waste interaction will however be maintained, to develop experimental methodologies in collaboration with other countries, and as a potential support to the decision making in those countries where a clay based near field is still the reference. The paper explains the current R&D approach, and highlights some recent experimental set-ups available at SCK•CEN for this purpose, with some illustrating results.Copyright © 2007 by ASME

One of the waste forms resulting from the Belgian programme for nuclear energy is the R7T7 high level waste glass, simulated by the SON68 glass. When exposed to distilled water or synthetic interstitial clay water, SON68 glass dissolves incongruently, releasing some elements (B, Li, Na) faster to the solution than others (Si, Al, Ca). The objective of this study is to describe the composition of the leachate in contact with inactive SON68 glass by assuming congruent glass dissolution followed by the precipitation of a secondary solid phase represented as an ideal solid solution in equilibrium with the leachate. Experimental SON68 dissolution data in distilled water was available at three temperatures. The solubility of the different end members in the solid solution is optimized using the available data for each temperature. The temperature dependence of the different end members was then obtained by Van 't Hoff equation. The calibrated model can describe the composition of the leachate and the altered glass phase during glass dissolution at different temperatures. The model, calibrated for the distilled water system, could successfully describe the composition of the leachate during glass dissolution experiments in a synthetic clay water system.

Integrated in situ corrosion tests with α-doped SON68 glass samples have been and are still being performed in the HADES underground laboratory in Mol, Belgium. Two test tubes have been dismantled, and the glass and clay samples have been analysed (mass loss, SEM, EDS, SIMS,…). The results on glass corrosion largely respond to the expectations and agree very well with results from surface laboratory tests and modelling predictions. The glass alteration is about two orders of magnitude higher at 90 °C than at 30 °C. The addition of powdered glass frit to a Ca-bentonite backfill reduces the glass alteration by two orders of magnitude in comparison with another backfill, dried Boom Clay. The α-doped glass samples were shown to have slightly thicker alteration layers. The thickness seemed to increase with increasing αβγ-activity. Because of the lack of sufficient data on mass loss, we cannot conclude that also the degree of alteration is higher for these samples. Under the thermal gradient in the test tubes, the reaction of Cabentonite with the glass samples and the powdered glass frit dispersed within the bentonite results in the neo-formation of non-swelling 7 Å minerals.

European Commission supported a wide research project entitled “Spent Fuel Stability under repository conditions” (SFS) within the 5th FWP, the aim of which was to develop a common understanding of the radionuclides release from spent nuclear fuel in geological disposal and build a RN release model in order to assess the fuel performance. This project achieved by the end of 2004 focuses both on the Instant Release Fraction (IRF) model and the Matrix Alteration Model (MAM).A new IRF model was developed based on the anticipated performances of the various fuel microstructures (gap, rim, grains boundaries) and the potential diffusion of RN before the canister breaching. However, this model lets the choice to the end-user about the degree of conservativeness to consider.In addition, fuel alteration has been demonstrated to be linked to the production of radiolytic oxidants by water radiolysis at the fuel interface, the oxidation of the fuel interface by radiolytic oxidants and the subsequent release of uranium under the influence of aqueous ligands. A large set of experimental data was therefore acquired in order (i) to upgrade the current radiolytic kinetic scheme, (ii) to experimentally correlate the fuel alteration rate and the fuel specific alpha activity by performing experiments on alpha doped samples, (iii) to experimentally test the potential inhibitor effect of hydrogen on fuel dissolution. Based on these results, a new MAM was developed, which was also calibrated using the experiments on inactive UO2 samples. This model was finally applied to representative granitic, salt and clayey environment to predict spent fuel long-term fuel performance.

Since reprocessing is no longer the reference policy in Belgium, studies on the direct disposal of spent fuel in a clay formation have gained increased interest in the last years. In order to determine to what extent the clay properties and the α-activity influence the dissolution kinetics of spent fuel for the long term disposal, static dissolution tests have been performed on 5 different types of α-doped UO2, representing a PWR fuel with a burn-up of 45 or 55 GWd · tHM−1 and fuel ages ranging between 150 and 90,000 years, in different Boom Clay (BC) media at room temperature and in an anoxic atmosphere for 90 to 720 days. The uranium activity in the clay fraction over time was found to be much higher than the U-activity in the leachates, which has been mainly ascribed to the high retention capacity of the BC. The average dissolution rate between 0 and 90 days obtained for the 5 types of α-doped UO2 were all found to be high and quite similar at ~263 µg · m−2 · d−1and a “longer-term” rate (181 to 720 days) ranging between zero and 15 µg · m−2· d−1. These results suggest that the activity of the fuels does not seem to have an effect on the UO2 dissolution rates under the considered test conditions. In order to study the partition/redistribution of U during UO2dissolution, sequential extraction experiments were performed. Results of the latter have provided a better mechanistic understanding of BC/spent fuel interaction processes as well as information on the role of the different minerals controlling the U-retention/immobilization.

Existing knowledge on glass dissolution and silica diffusion in Boom Clay is validated by experiments where both phenomena could be studied simultaneously. SON68 glass coupons, doped with radioactive 32Si, were sandwiched between two cores of fresh humid Boom Clay and heated to 30°C. At the end of the experiment, the system was dismantled, the mass loss of the glass coupon was measured, and the clay core was sliced to determine the diffusion profile of the 32Si dissolved from the glass. These data were completed with analyses of the clay water and surface analyses for analogous tests with undoped glass. The results are interpreted by assuming congruent glass dissolution at a constant rate, with a glass silica saturation concentration between 14 and 20 mg/l, a forward glass dissolution rate (at zero silica concentration) of 0.028 g.m−2day−1, an apparent silica diffusion coefficient in the clay of 1.4 10−12 m2sec−1, and a distribution coefficient for silica on Boom Clay between 0.010 and 0.075 m3kg−1. These parameter values are close to the range found in literature. It was not necessary to consider diffusion through the gel, precipitation or detailed geochemical reactions. The modeling exercise shows that the existing knowledge about the subsystems glass and clay can succesfully be integrated to describe the coupled processes in the whole system.

This paper reviews the main conclusions of two European Commission funded projects, GLASTAB and GLAMOR, on the durability of HLW glass in geological disposal, and the remaining uncertainties. The progress in GLASTAB relates to the characterization of the alteration layer, the modeling of glass dissolution, the interaction of glass with near field materials, the behaviour of radionuclides, and the calculation of HLW glass dissolution as part of geological disposal. The GLAMOR programme was focused on the decrease of the glass dissolution rate upon achieving silica saturation in solution. Two conclusions from this programme are the importance of the residual rate measured beyond silica saturation and the need to reduce the uncertainties in model parameters. The overall conclusion from the two projects is that strong progress has been achieved on basic dissolution mechanisms and interaction with disposal environments - HLW glass will act as a good barrier to the release of radionuclides in geologic disposal.

The solubility values usually used in Belgium for Performance Assessment of the disposal of HLW in deep geological formations are based on migration experiments in the far-field and, thus, can be considered as "far-field solubilities". The particular chemistry in the near-field of the repository will also influence the solubility of the released radionuclides. Therefore, the determination of maximum concentrations of the radionuclides in the near-field is also useful. This was approached by leach tests with doped glass in near-field conditions. The experiment consisted of static leach tests with the french R7T7 reference glass doped with 237Np (0.34 wt%) and 99Tc (0.12 wt%), two key radionuclides in the performance assessment of the disposal system. The glass was put in contact with FoCa7-clay together with pyrite, metallic iron and Boom clay water to simulate the near-field conditions under reducing conditions. The interpretation of the results is based on the measurement of the steady-state Np and Tc concentrations in the leachates after ultrafiltration through 100,000 MWCO membranes and on the activity measured on the clay after different test durations. Most of the Np and Tc activity (99.5%) was found associated to the clay and increasing as a function of time while the Np and Tc concentration in the leachates were stable and extremely low with a mean value of 1.5×10−10M for Np and 5.6×10−10M for Tc. This is due at least partially to sorption on the clay, but there are indications for precipitation as well.

To estimate the lifetime of vitrified high level waste (HLW-glass) in Boom Clay disposal conditions, the dissolution behaviour of waste glass has been studied with experiments performed in surface laboratories and in the HADES underground research facility of SCK·CEN since 1980. We present the main topics and first results of the SCK·CEN programme 2000–2003. This programme focuses on the following items: (1) the diffusion/sorption/precipitation of silica in Boom clay or backfill clay, (2) demonstration of glass dissolution behaviour in realistic test conditions, (3) the effect of presaturation of the clay with silica, and (4) the estimation of near field concentrations of critical isotopes. The experiments have shown so far that Si, released by the glass, is effectively immobilized by Boom Clay, but it can nevertheless diffuse into the clay without immediately precipitating. The dissolution rate of glass SON68 and SM539 is determined in Boom Clay at in situ density and at 30°C (this is the long-term temperature expected near the waste glass packages in a Boom Clay repository). The dissolution rates, based on glass mass losses, are constant during the first year, at ∼ 0.010 g.m−2 .day−1 for glass SON68 and ∼ 0.012 g.m−2 .day−1 for glass SM539. The addition of glass frit causes a decrease of the glass dissolution rate, both with glass SON68 and SM539, and both in Boom Clay and in FoCa-clay. In FoCa-clay at high density with glass frit, the dissolution rates, based on glass mass losses, after 8 months at 30°C are ∼ 0.001 g.m−2 .day−1 (SM539) and ∼0.005 g.m−2 .day−1 (SON68). Because the experiments performed in Boom Clay and FoCa-clay with glass frit simulate realistic conditions (high clay density, low temperature), they can be used to estimate the maximum glass dissolution rate in a (Boom) clay repository. The corresponding minimum lifetime of a glass canister, calculated with the SCK·CEN code for lifetime predictions, is of the order of 105 to 106 years, if we neglect the internal glass surface area (due to cracking). In more diluted clay suspensions with glass frit, the glass dissolution rate is 10−4 to 10−5 g.m−2 .day−1 or even zero. This would correspond to a lifetime of >>106 years. So far, there is no indication that the addition of glass frit leads to secondary phase formation at low temperature (30–40°C). Leach experiments with doped glasses SON68 and SM539 suggest that the maximum concentrations of most critical radionuclides in near field conditions are lower than the best estimate solubilities used for performance assessment studies in Boom Clay. For Se, relatively high concentrations were measured, though. The research programme for the underground laboratory is not discussed.Copyright © 2003 by ASME

An analytical model is developed to derive diffusion coefficients from the element profiles of the most mobile glass elements. Congruent dissolution is considered as a transient phenomenon and neglected. In the model, altered glass consists of two sublayers: a gel layer at the solution side, and a diffusion layer at the pristine glass side. In both layers, mobile glass elements diffuse towards the solution, while hydrogen/water diffuses towards the pristine glass.Fitting the element profiles of boron, sodium and lithium leads to diffusion coefficients in the diffusion layer, which at 90°C are typically of the order of 10-22m2/s to 10-21m2/s. At 40°C, these diffusion coefficients are typically a factor ten lower. The diffusion coefficients in the gel are by less than a factor two hundred larger than those in the diffusion layer.The diffusion coefficients derived from the profiles in this way are used to predict the dissolution behavior. These predictions, for which a constant value (representing congruent dissolution) is added to the predicted diffusive release, agree reasonably well with the measured dissolution data. Those results support the idea that diffusion determines the long term release of glass matrix elements in synthetic interstitial clay water. Flux conservation at the interface between the gel and the diffusion layer is not consistent with a constant (as a function of time) width of the diffusion layer. Some experimental evidence supports that this width increases with time.

Boom Clay is the candidate geologic formation in Belgium for disposal of vitrified high level waste. In the waste glass, 237Np and 99Tc are some of the principal radionuclides. The leaching behaviour of these nuclides has been studied in dissolution tests involving glass doped with 237Np and 99Tc and was found to depend strongly on the redox conditions of the media contacting the glass. Static tests with the reference R7T7 glass and the reference PAMELA SM539 glass were performed in two clay media that may interact with the glass during the geological disposal. The first test medium consisted of FoCa-clay (Fourges-Cahaignes-clay), which is an oxidised clay. The second test medium was a mixture of backfill material, consisting mainly of FoCa-Clay, with metal corrosion products and Boom Clay. Lower Np and Tc concentrations were found in Boom Clay compared to FoCa-clay. In FoCaclay, Np and Tc were predominately in their oxidised form. Although Tc is mostly present in the soluble pertechnetate form, the Np concentrations are lower than expected for the chemistry of Np(V) in the medium. The concentrations of the radionuclides are steady-state concentrations rather than thermodynamic concentrations. Different mechanisms can be invoked to explain these Np concentrations. In Boom Clay, the measured Np concentrations are close to the solubility limit whereas the Tc concentrations are slightly higher than those calculated in Boom Clay at thermodynamic equilibrium. The initial specific activity of the radionuclide in the glass was found to influence the soluble concentration. Np and Tc were not found to be retained in the glass reaction layer for both media.

The effect of γ-radiation on the dissolution of candidate high-level waste glasses was investigated in potential disposal environments at 90°C. The media consisted of mixtures of Boom Clay, bentonite clay and cement with clay water. During the experiments the pH of Boom Clay decreased, probably mainly by radiolytical oxidation of pyrite. The addition of bentonite, cement and glass buffered the pH decrease. Under radiation the glass mass losses decreased, whereas the leach rate of soluble elements was not influenced or appeared to increase. This is explained through the radiolytical acidification, and possibly by bubble formation in the glass.

Static dissolution experiments were performed with unirradiatcd UO2 in Boom Clay water. The objectives were (1) to measure the solubility of uranium species in Boom Clay water, with UO2 as the solid phase, and (2) to assess the impact of dissolved organic matter and carbonate concentration on this solubility. The tests were supported by calculations with geochemical codes to indicate possibly solubility controlling solid phases. The tests were performed in anoxic and reducing conditions, at 20 and 25°C. The following conclusions could be drawn: (1) Within 2 months in anoxic conditions, the uranium concentrations appear to approach saturation. (2) The “near-saturation” concentrations are between 2.4 and 7.8× 10−7 M. (3) The influence of the carbonate concentration and humic acids on the uranium concentration was apparently small, but the interpretation is hampered by pH and Eh variations. (4) The concentrations tend to be higher than the concentrations found in literature for similar Eh and/or pH conditions; this can probably be explained by small differences in experimental conditions. (5) The measured “near steadystate” uranium concentration in the real clay water agrees relatively well with the solubility calculated for uraninite. (6) Addition of sulfide species reduced the rcdox potential, but not the uranium concentrations, except in real Boom Clay water.

A geochemical model has been developed to describe the interactions of the main elements released from the COGEMA R7T7 SON68 glass with the Boom Clay environment and predict the formation of secondary phases which may influence the glass corrosion. It is based on two submodels: a model describing the interaction between the clay and its porewater, and a model for glass corrosion in groundwater. According to the calculations, no secondary minerals are predicted to precipitate when glass dissolves in the Boom Clay. The oncentrations of the main elements released from the glass are controlled by stable phases present in the Boom Clay (a.o. Si, Al) or by cation exchange reactions (e.g. Na). The concentration of silica in solution can be maintained below the glass silicon solubility limit. Therefore, the Boom Clay can suppress the saturation of silica, explaining the enhancing effect of Boom Clay on glass corrosion. The influence of the presence of clay on the other components of the glass is also discussed.

The leaching behaviour of Pu, Am, Np and Tc from the high-level waste glasses of interest to Belgium has been studied. Corrosion tests were carried out on inactive glasses (the Cogéma R7T7 glass SON68 and the DWK Pamela glasses SM513 and SM527), doped with the radionuclides. Nuclide leaching was studied in contact with a slurry of Boom clay, diluted with claywater, at 90°C and 40°C, for durations of up to 4 years. The retention of Pu and Np in the glass surface layers was considerable for SON68 and SM513, but small for SM527. The retention of these elements decreased at lower temperature. Tc was not retained. The largest part of the leached activity is adsorbed onto the clay in an immobile form. Nuclide concentration in mobile form (smaller than 105 MWU) was found to be independent on duration for Pu and Am. Increasing concentrations were found for Np and Tc. The mobile concentrations were smaller in the presence of corrosion products.

The Al2O3 rich borosilicate glass SM527 was submitted to corrosion tests with glass surface area to solution volume ratios ranging from 10 to 10000 m-1. This latter condition would correspond with a 1000 fold acceleration relative to the reference MCCI condition. Powdered glass was used to reach SANV ratios of 500 m-1 and more. The leaching solutions were either distilled water or referred to Boom clay disposal conditions. The results based on the boron concentration in solution revealed a relatively linear dependence on SA/V.t0.5 in the pure solutions, on the longer term (DW and clay water). Diffusion is suggested to be the process governing the glass dissolution, although other processes should not be excluded. In a clay / clay water mixture (slurry), long term dissolution seems to be limited by saturation. Short term data for boron are largest in the clay slurry, but with time the boron concentrations converge to similar values in the three media considered. The use of SA/V as an accelerating factor is promising, but certainly requires additional research.

Leaching experiments with deionised water at 70°C have been carried out to ascertain the effect of oxidising and anaerobic conditions and the presence of Boom clay on Np release from Synroc.The normalised solution leach rates of Np measured in these experiments are about a factor of 10 lower under reducing conditions than oxidising conditions and the presence of Boom clay was found not to enhance the leaching of Np from Synroc. Neptunium was found to be predominantly in the mobile fraction and was not removed from the leaching solutions significantly by filtration or ultrafiltration.

A consortium of 10 partners from seven European countries and the European Commission collaborated in investigating the short-term release of radionuclides from disposed spent nuclear fuel upon canister failure. The Collaborative Project FIRST-Nuclides was implemented in the scope of the 7th Euratom Framework Programme in the period from 2012 to 2014. The objectives and organization of the project are presented, as well as the experiments with highly radioactive materials under investigation. The outcome of the project summarizes the measured instant release fraction (IRF) of safety-relevant isotopes from high burnup spent UO2 nuclear fuels (SNFs). Specifically discussed are the dependencies of the IRF on the sample properties, the gap and grain boundary releases, and the behavior and IRFs of elements such as cesium, iodine, and selenium. The IRFs of nonstandard SNFs were also investigated. The summary is complemented by the presentation of the modeling approaches within the project.

The vitrification of nuclear waste salts similar to the salts present in Eurobitum produced by the former Eurochemic reprocessing plant, Mol-Dessel, Belgium, is investigated. In order to approach the composition of the high-level waste reference glass SON68, mixtures with different ratios of glass frit to waste salts have been prepared and vitrified. The concentration of the waste salts varied between 17 and 25 % by weight.To compare the short term corrosion behaviour with reference waste glass, leach tests have been performed. During 8 weeks, the glass samples were submitted to leach tests at 363 K using Milli Q water as leachant, at a glass exposed area over leachant volume ratio equal to 100 m−1.During this short period, scanning electron microscopy with energy dispersive X-ray analysis (SEM-EDX) showed no gel layer formation. This technique did not detect changes in the chemical composition. However, X-ray photoelectron spectroscopy (XPS) showed that the binding energy of the Zr 3d and Si 2p electrons decreased as a function of leaching time, indicating an increasing covalent character of the glass surface. Inductively coupled plasma-mass spectrometry (ICP-MS) of the leachate as a function of time shows that ionic species such as Na and K leach faster than e.g. Si and B. These results together with the XPS data suggest the onset of a gel layer formation. The effect of waste content in the glass on the leaching rate is also studied. The results are compared with similar leaching tests on R7T7 type SON68 glass. Most elements are better retained in the simulated Eurobitum glass than in the SON68 glass.

This study is focused on the alteration behavior of spent nuclear fuel when exposed to highly alkaline groundwater. Contact of highly alkaline solution with the waste product is considered in the Belgian concept for disposal in the Boom Clay formation. According to the “supercontainer design” the fuel will be encapsulated in carbon steel canisters, surrounded by a concrete over-pack. After saturation of the engineered barriers by porewater, interactions with the concrete will result in solutions rich in NaOH, KOH and Ca(OH)2. Using this type of solution at pH 12.5, spent nuclear fuel corrosion experiments were conducted over 378 days. Under anoxic conditions, parallel experiments were performed (a) in the absence of Fe phases and (b) in the presence of solid Fe phases representing container (corrosion) products. Both types of experiments resulted in relatively low matrix dissolution rates, around 10-7 per day, according to the fractional release of Sr. Solution concentrations of actinides are close to or below the detection limit, indicating an effective retention of these radioelements in the system. The observed precipitation of a Ca rich phase onto the surfaces of the corroded fuel samples may be related to the inhibited re-lease of actinides, Sr and other matrix bound radioelements.