Andreas Loida

Anoxic corrosion of various high burnup spent fuel samples

Abstract To assess the long-term performance of spent fuel in a repository in saline environments, the effect of iron and sample dimension on the corrosion behavior and radionuclide release from high burnup fuel was studied. Additionally, the release contributions of the highly burned fuel rim and of α-recoil transport mechanism was evaluated. Results show that spent fuel is thermodynamically unstable even under reducing conditions (presence of Fe) but the dissolution rates are very slow. Slow dissolution rates are also encountered with powdered fuel, probably resulting from depletion of radiolytic reactants.

Processes controlling radionuclide release from spent fuel

Dissolution of spent fuel has been studied in saline, anaerobe, carbonate free solutions. Processes controlling spent fuel dissolution and associated radionuclide release are radiolytically controlled oxidative dissolution, sorption on container, solubility and coprecipitation. Upper limits for oxidative dissolution rates are given by the production rates of oxidative radiolysis products. This limitation leads to a strong decrease in surface area normalized reaction rates with increasing surface to volume ratio (S/V) and imposes geometric constraints on prediction of spent fuel behavior in a repository. Solution concentrations of Am during spent fuel corrosion were about 5 orders of magnitude lower than the solubility of Am(OH){sub 3}(s) and are likely controlled by coprecipitation. Pu concentrations may be controlled by Pu(VI) or Pu(IV) (hydr)oxides.

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.

Coprecipitation phenomena during spent fuel dissolution. Part 1: Experimental procedure and initial results on trivalent ion behaviour

Coprecipitation may be a significant process in controlling radionuclide release during spent fuel dissolution in geological disposal. The coprecipitation behaviour of the trivalent elements Am, Cm and Eu with Na-polyuranates is studied in saline solutions. Experimental data are obtained by precipitation from a supersaturated solution of dissolved high burnup spent fuel in 5 m NaCl solution at pH values between 5.7 to 12. The resulting concentrations of dissolved Am, Cm and Eu were found similar to those observed in spent fuel leaching experiments and were found to be considerably lower than expected, if the formation of pure Am(OH)3, Cm(OH)3 and Eu(OH)3 solid phases is assumed. Coprecipitation phenomena and formation of solid solutions are suggested to be responsible for the lower solubility concentrations.

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.

Long-term stability of spent nuclear fuel waste packages in Gorleben salt repository environments

The results are summarized of 15 yr of German research on spent el with respect to its suitability as a waste form disposed of in a repository located in the Gorleben salt dome. Within the multibarrier system for long-term isolation of high-level waste (HLW), the innermost engineered barrier canistered spent el contributes essentially to isolating radionuclides from the biosphere if a salt brine were to come into contact with the waste form. A large fraction of the radionuclide contents of the reacted jkel mass would become reimmobilized within secondary alteration products and on container corrosion products, but inevitably a certain nuclide-specific fraction would be released into the aqueous geochemical environment. The corrosion resistance of the fuel and the radionuclide mobility are not inherent materials properties but also depend on geological disposal conditions, packing concepts, and radioactive decay. In particular the availability of oxidants is critical, controlling spent-fuel alteration rates and alteration products as well as radionuclide solubilities. Spent fuel is at least as suitable for final disposal as is HLW glass.

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.

Radionuclide Release From Spent MOX Fuel

In Germany certain quantities of spent MOX fuel will be disposed in a rock salt repository. In order to assess the safety of the repository the corrosion behavior of spent MOX fuel in salt solution must be understood. Since material properties of spent MOX fuel are different compared to these of spent LWR UO 2 fuel, it is not possible to deduce from corrosion behavior of the latter to that one of spent MOX fuel. It is expected that the most significant differences will occur in the dissolution kinetics, whereas similar solid alteration products are expected from the similar overall composition of the fuels. Various experiments were performed focused on the effect of fuel type on the dissolution kinetics. The results of this work are summarized and evaluated

Mobility and Criticality of Plutonium in a Repository

Abstract In an underground repository for spent fuel, criticality is excluded initially by compliance with the disposal conditions. In the long term, critical accumulations of fissile material can be formed only by mobilization of uranium and plutonium from the waste forms and subsequent precipitation or sorption of these elements. This paper presents an overview of mechanisms relevant for mobilization and possible accumulation of U and Pu from disposed mixed-oxide fuel elements. Concentrations of fissile materials observed in laboratory corrosion experiments together with model approaches are applied to determine the degree of fissile material accumulation and the risk of a sustained nuclear chain reaction. A prerequisite of criticality in a repository is an accumulation of fissile materials. Since geometry, moderation, and neutron absorption properties cannot be forecast, the neutron multiplication factor kinf is used (instead of keff) as a measure of the incidence of criticality. The factor kinf is derived for several scenarios. Required critical masses and critical volumes are evaluated. The accumulation of Pu onto solids is considered, and it is shown how selective enrichment of Pu and U may affect the risk of criticality. It is also shown that the criterion for criticality would be met only in the unrealistic case of selective sorption of 239Pu. Realistic sorption densities are too low to provide sufficient accumulation of fissile materials for criticality. This is particularly true if high Cl concentrations are present.

Impact of Hydrogen Overpressure on Spent Fuel Alteration Behavior in Salt Brine During a Three-Year Long Corrosion Test

The corrosion of the Fe-based spent fuel containing canister produces large amounts of hydrogen, which dominate the composition of the gas phase. To quantify to what extent the hydrogen over- pressure may counteract radiolysis enhanced matrix dissolution, related experimental work has been performed. High burnup spent fuel was corroded in 5 M NaCl solution in different tests: (1) Fe ab-sence ( 2 overpressure), (2) Fe presence (2.8 bar H 2 overpressure), (3) Fe absence, external applied H 2 overpressure of 3.2 bar. In the absence of Fe, after the application of H 2 overpressure the UO 2 matrix dissolution rate decreased by a factor of about 10. The concentrations of U, Np, Tc in so- lution were found to be decreasing at least two orders, and ranging within the same level as in the presence of Fe powder. However, Pu and Am concentrations were less affected, due to the absence of Fe powder and the associated high sorption capacity for these radioelements.