Volker Metz

The effect of pH and temperature on kaolinite dissolution rate under acidic conditions

Abstract The main goal of this paper is to demonstrate a new rate law describing the combined effect of pH (0.5 to 4.5) and temperature (25°C to 70°C) on kaolinite dissolution rate, under far from equilibrium conditions, as a step towards establishing the full rate law of kaolinite dissolution under acidic conditions. Dissolution experiments were carried out using non-stirred flow-through reactors fully immersed in a thermostatic water bath held at a constant temperature of 25.0°C, 50.0°C or 70.0°C ± 0.1°C. Kaolinite dissolution rates were obtained based on the release of silicon and aluminum at steady state. The results show good agreement between these two estimates of kaolinite dissolution rate. Kaolinite dissolution rates range as a function of temperature and fluid composition from 8 ± 1 × 10 −15 mol m −2 s −1 (at 25°C and pH 4.5) to 1.5 ± 0.2 × 10 −11 mol m −2 s −1 (at 70°C and pH 0.5). In general, dissolution rate increases with temperature and decreases with pH. The combined effect of pH and temperature is modeled by two independent proton promoted reaction paths. The first reaction path controls the overall dissolution rate at pH ≥ 2.5, whereas the second path controls it below pH 0.5. Between pH 0.5 and 2.5 the two reaction paths influence the rate. Using this model the effects of pH and temperature on the overall dissolution rate of kaolinite under acidic condition can be described by: Rate=2·10 2 ·e −22/RT · 2·10 −10 ·e 19/RT ·a H + 1+2·10 −10 ·e 19/RT ·a H + +5·10 7 ·e −28/RT · 1.4·10 −7 ·e 10/RT ·a H + 1+1.4·10 −7 ·e 10/RT ·a H + where R is the gas constant, T is the temperature (K) and a H + is the activity of protons in solution.

Stirring effect on kaolinite dissolution rate

Abstract Experiments were carried out measuring kaolinite dissolution rates using stirred and nonstirred flow-through reactors at pHs 2 to 4 and temperatures of 25°C, 50°C, and 70°C. The results show an increase of kaolinite dissolution rate with increasing stirring speed. The stirring effect is reversible, i.e., as the stirring slows down the dissolution rate decreases. The effect of stirring speed on kaolinite dissolution rate is higher at 25°C than at 50°C and 70°C and at pH 4 than at pHs 2 and 3. It is suggested that fine kaolinite particles are formed as a result of stirring-induced spalling or abrasion of kaolinite. These very fine particles have an increased ratio of reactive surface area to specific surface area, which results in enhancement of kaolinite dissolution rate. A balance between production and dissolution of the fine particles explains both the reversibility and the temperature and pH dependence of the stirring effect. Since the stirring effect on kaolinite dissolution rate varies with temperature and pH, measurement of kinetic parameters such as activation energy may be influenced by stirring. Therefore, standard use of nonagitated reaction vessels for kinetic experiments of mineral dissolution and precipitation is recommended, at least for slow reactions that are surface controlled.

The INE-Beamline for actinide science at ANKA

Since its inauguration in 2005, the INE-Beamline for actinide research at the synchrotron source ANKA (KIT North Campus) provides dedicated instrumentation for x-ray spectroscopic characterization of actinide samples and other radioactive materials. R&D work at the beamline focuses on various aspects of nuclear waste disposal within INEs mission to provide the scientific basis for assessing long-term safety of a final nuclear waste repository. The INE-Beamline is accessible for the actinide and radiochemistry community through the ANKA proposal system and the European Union Integrated Infrastructure Initiative ACTINET-I3. Experiments with activities up to 1 × 10(+6) times the European exemption limit are feasible within a safe but flexible containment concept. Measurements with monochromatic radiation are performed at photon energies varying between ~2.1 keV (P K-edge) and ~25 keV (Pd K-edge), including the lanthanide L-edges and the actinide M- and L3-edges up to Cf. The close proximity of the INE-Beamline to INE controlled area labs offers infrastructure unique in Europe for the spectroscopic and microscopic characterization of actinide samples. The modular beamline design enables sufficient flexibility to adapt sample environments and detection systems to many scientific questions. The well-established bulk techniques x-ray absorption fine structure (XAFS) spectroscopy in transmission and fluorescence mode have been augmented by advanced methods using a microfocused beam, including (confocal) XAFS/x-ray fluorescence detection and a combination of (micro-)XAFS and (micro-)x-ray diffraction. Additional instrumentation for high energy-resolution x-ray emission spectroscopy has been successfully developed and tested.

Surface protonation data of kaolinite—reevaluation based on dissolution experiments

The aim of the present study is to compare available surface titration curves of kaolinite, to explain the differences between them, and to constrain their interpretation based on predictions of surface protonation that emerged from dissolution experiments. Comparison of six surface titration curves obtained at 25 degrees C reveals significant discrepancies, both in the shape of the curves and in the pH of the point of zero net proton charge (pH(PZNPC)). Based on an analysis of the different sites available for adsorption on kaolinite surfaces we conclude that different kaolinite samples are expected to have similar pH(PZNPC). Therefore, the major reason for the differences in the observed surface protonation is related to the different ways in which the pH(PZNPC) was determined. To compare the titration curves, some of the curves were recalculated so that the proton surface concentrations of all the titration curves would be zero around pH 5. As a result, we obtained a good agreement between the titration curves. A prediction of the molar fraction of protonated sites was retrieved from modeling of kaolinite dissolution reaction and was compared to the protonation data obtained from surface titration. The model successfully predicts the surface protonation data of most of the surface titrations.

Radionuclide behaviour in the near-field of a geological repository for spent nuclear fuel

Abstract Even though chemical processes related to the corrosion of spent nuclear fuel in a deep geological repository are of complex nature, knowledge on underlying mechanisms has very much improved over the last years. As a major result of numerous studies it turns out that alteration of irradiated fuel is significantly inhibited under the strongly reducing conditions induced by container corrosion and consecutive H2 production. In contrast to earlier results, radiolysis driven fuel corrosion and oxidative dissolution appears to be less relevant for most repository concepts. The protective hydrogen effect on corrosion of irradiated fuel has been evidenced in many experiments. Still, open questions remain related to the exact mechanism and the impact of potentially interfering naturally occurring groundwater trace components. Container corrosion products are known to offer considerable reactive surface area in addition to engineered buffer and backfill material. In combination, waste form, container corrosion products and backfill material represent strong barriers for radionuclide retention and retardation and thus attenuate radionuclide release from the repository near-field.

Effects of hydrogen and bromide on the corrosion of spent nuclear fuel and γ-irradiated UO2(s) in NaCl brine

Abstract Radiation induced UO2(s) corrosion is studied at elevated hydrogen pressure in NaCl brine containing traces of bromide. Release of Sr, Cs, Tc and actinides was measured in corrosion experiments with spent nuclear fuel pellets in presence of 10−2 mol H2 (kg H2O)−1, and 10−4 and 10−3 mol Br− (kg H2O)−1, respectively. For comparison, depleted UO2(s) pellets were γ-irradiated in NaCl brine at 10−3 mol H2 (kg H2O)−1 and 0−10−4 mol Br− (kg H2O)−1, respectively. In the γ-radiolysis experiments a significant increase in the yield of radiolytic products due to Br− is observed. Both, in the γ-radiolysis experiment with Br− and in that without Br−, the UO2(s) sample was oxidized, and the concentration of dissolved uranium was controlled by precipitation of meta-schoepite and clarkeite. In the spent nuclear fuel corrosion experiment under H2 overpressure, aqueous concentrations of Tc and Np were in the range of solubilities of Tc(IV) and Np(IV) hydroxides, whereas measured U concentrations were between solubilities of U(VI) and U(IV) phases. The release rate of Sr was significantly increased in the presence of Br− traces. Results of the complementary spent nuclear fuel corrosion and γ-radiolysis experiments allow the conclusion that Br− traces reduce significantly the protective hydrogen effect with respect to the release of certain radionuclides and the yield of radiolytic products.

Chemical status of U(VI) in cemented waste forms under saline conditions

Abstract Retention of U(VI) in cemented waste forms reacting with NaCl and MgCl2 brines is investigated in long-term leaching experiments on full scale monoliths. Solution compositions were monitored over a period of 17 to 18 years. After termination of the leaching experiments, chemical and mineralogical compositions of solid reaction products were studied intensively. XRD, TRLFS and XANES/EXAFS analyses indicate uranophane (Ca(UO2)2(SiO3OH)2·5H2O) to be the dominant uranium bearing phase in the corroded cement. Other possible uranium phases such as soddyite, meta-schoepite, and di-uranate phases could not be identified by combining the results of the various experimental techniques.

Geochemically derived non-gaseous radionuclide source term for the Asse salt mine: assessment for the use of a Mg(OH)2-based backfill material

Summary The Asse salt mine was used as a test site for radioactive waste disposal from 1967 to 1978. Low- and intermediate-level waste forms (LLW/ILW) were emplaced, containing a total radionuclide inventory of 3×1015 Bq, estimated for reference date January 1, 2005. It is expected that MgCl2-rich brine will enter the emplacement rooms and react with the cemented waste products. Possible microbal degradation of organic waste components in MgCl2-rich brine could produce significant quantities of CO2, resulting in an acidification and consequently in an increase of element solubilities. Application of buffering backfill materials is discussed for closure of the mine. Selection of backfill materials is based on geochemical modeling taking into account the corrosion of the cemented LLW/ILW and degradation of organic waste components. In the present study the evolving geochemical milieu and respective solubilities of Am, Np, Pu, U, Th, Tc, Sr, Cs and I were modeled for each emplacement room. Laboratory experiments were undertaken to selectively verify the modeling predictions. Geochemical modeling leads to the conclusion that Portland cement, a Mg(OH)2-based material and crushed salt can be used in different combinations as backfill materials. According to the predictions Mg(OH)2-based backfill material controls the pH and concentration of dissolved inorganic carbon within ranges that are favorable with respect to actinide solubilities. A source term for radioelements and fission products is derived from a comparison of the respective solubilities and the radionuclide inventories. Calculated solubilities of Th, Pu and U are lower than their inventories. These are controlled by the thermodynamic solubility therefore. Maximum solution concentrations of all other elements are controlled by their inventory.

Geochemically Based Safety Assessment

Performance assessment (PA) for nuclear waste disposal based on transport processes frequently neglects significant safety factors by overestimation of radionuclide mobilization and underestimation of radionuclide retention processes. To include well understood geochemical knowledge into PA, the quasi-closed system approach (QCS) was developed. The QCS approach is described and applied to a LLW disposal in a German salt mine with respect to the disposed waste forms, geo-engineered barriers, and backfill strategies. The geochemical tools and the thermodynamic database for modelling highly concentrated salt systems are discussed. Applications are demonstrated which cover the long-term geochemical environment in the disposal caverns, the optimization of buffer materials, radionuclide retention, and the overall robustness of the approach. Also the effect of a potential solution exchange between different emplacement caverns is investigated. It is shown that the QCS approach provides essential data concerning the long-term geochemistry and related radionuclide concentrations to be used in PA and safety analysis.

Corrosion Behavior of High Burnup Spent Fuel in Highly Alkaline Solutions

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.