Bernhard Kienzler

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.

Long-Term Leaching Experiments of Full-Scale Cemented Waste Forms: Experiments and Modeling

Abstract Experimental findings of full-scale leach tests performed on simulated cemented waste forms and self-shielded concrete waste containers for periods up to 19 yr in saturated salt brines (NaCl- and Q-brine) are presented. Measurements cover the evolution of leachant composition and the release of radionuclides such as Cs, U, and Np. Performance of the waste forms and the self-shielded concrete waste containers depends on the pore volume of the hardened cement/concrete, which is correlated to the water/cement ratio of the waste forms. Cesium release follows a linear time dependence. Samples, especially those having a high pore volume, show almost complete release of Cs in the period of investigation. Uranium release is independent of the leach period. Uranium concentrations are controlled by thermodynamic equilibrium. Neptunium is released only to a small extent; concentrations are close to the detection limit. Modeling of the cement corrosion progress allows the prediction of the evolution of the brines in terms of pH, calcium concentration, etc. and the identification of solids controlling the solubilities of the main components and of uranium.

Swedish-German actinide migration experiment at ASPO hard rock laboratory.

Within the scope of a bilateral cooperation between Svensk Kärnbränslehantering (SKB) and Forschungszentrum Karlsruhe, Institut für Nukleare Entsorgung (FZK-INE), an actinide migration experiment is currently being performed at the Aspö Hard Rock Laboratory (HRL) in Sweden. This paper covers laboratory and in situ investigations on actinide migration in single-fractured granite core samples. For the in situ experiment, the CHEMLAB 2 probe developed by SKB was used. The experimental setup as well as the breakthrough of inert tracers and of the actinides Am, Np and Pu are presented. The breakthrough curves of inert tracers were analyzed to determine hydraulic properties of the fractured samples. Postmortem analyses of the solid samples were performed to characterize the flow path and the sorbed actinides. After cutting the cores, the abraded material was analyzed with respect to sorbed actinides. The slices were scanned optically to visualize the flow path. Effective volumes and inner surface areas were measured. In the experiments, only breakthrough of Np(V) was observed. In each experiment, the recovery of Np(V) was < or = 40%. Breakthrough of Am(III) and Pu(IV) as well as of Np(IV) was not observed.

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.

ACTINIDE MIGRATION IN FRACTURES OF GRANITE HOST ROCK : LABORATORY AND IN SITU INVESTIGATIONS

Abstract Within the scope of a cooperation between Svensk Kärnbränslehantering AB and Forschungszentrum Karlsruhe, Institut für Nukleare Entsorgung, a series of actinide migration experiments were performed both in the laboratory and at the Äspö Hard Rock Laboratory in Sweden. The objectives of these experiments were to quantify the sorption of different actinide elements in single fractures of a granite host rock and to investigate the sorption mechanisms. To guarantee the most realistic conditions—as close to nature as possible—in situ experiments were performed in the Chemlab 2 borehole probe. These migration experiments were complemented by laboratory sorption and migration studies. The latter included batch experiments with flat chips of natural material extracted from fracture surfaces to identify the mineral phases relevant to radionuclide sorption by means of autoradiography. Scanning electron microscopy analyses provided information on the composition of sorption-relevant phases and X-ray photoelectron spectroscopy of Np, Tc, and Fe distribution revealed the redox states of these elements. Important mineral phases retaining all actinides and Tc were Fe-bearing phases. From the migration experiments, elution curves of the inert tracer (HTO), Np(V), U(VI), and to a small extent of Tc(VII) were obtained. Americium(III) and plutonium(IV) were not eluted. The mechanisms influencing the migration of the elements Np, U, and Tc depended on redox reactions. It was shown by various independent methods that Np(V) was reduced to the tetravalent state on the fracture surfaces, thus resulting in a pronounced dependence of the recovery on the residence time. Technetium was also retained in the tetravalent state. Elution of natural uranium from the granite drill cores was significant and is discussed in detail.

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.

Speciation of Actinides in Granite Subjected to Tracer Studies

Tomographic studies of granite from the Aspo Underground Laboratory in Sweden are performed to understand the observed release of natural uranium in column studies upon application of groundwater flow. X-ray phase-sensitive tomography images reconstructed from in-line X-ray phase contrasted radiographs were compared with scanning fluorescence tomography reconstructions. The latter are based on sinograms of X-ray emission lines recorded with spatial resolution on a nanometer scale for a granite rock containing U after being subjected to a radionuclide tracer experiment. The results show that the U distribution follows microfissures or pores in the granite. Naturally occurring Th is revealed to be intimately associated with what appears to be feldspar, suggesting its being present as a dopant ion in the mineral structure. Neptunium tracer was not found in a sample prepared using FIB lift-out techniques, although the presence of Np was identified in the original granite piece during screening experiments prior to FIB-ing. The Np-containing part of the sample broke off during the preparation. Although this case study was limited to only three samples, initial interpretation of results confirm that Th is less likely than U to become mobile in groundwater upon drilling and excavation of emplacement caverns and shafts for a nuclear waste repository in granite bedrock.

Corrosion Behavior of Pre-oxidized High Burnup Spent Fuel in Salt Brine

During long-term interim storage of spent fuel, pre-oxidation of the UO 2 -matrix may not be ruled out completely. This can happen if air could find access to the fuel in the case of cladding failure. The aim of this work is to study the impact of pre-oxidation of the fuel surface on the UO2 matrix dissolution rate and the associated mobilization or retention of radionuclides in highly concentrated salt solutions. The tests were performed with samples that suffered pre-oxidation during up to seven years. The dissolution rate of a fuel sample contacted by small quantities of air-oxygen was found to be roughly a factor of 10 higher in comparison to non oxidized samples, but concentrations of radionuclides, especially Pu and U were hardly affected. The majority of dissolved radionuclides, especially Pu, U appear to have been reimmobilized on the fuel sample itself.