The Influence of Near Field Redox Conditions on Spent Fuel Leaching

UO2 dissolution in bicarbonate solution with H2O2: the effect of temperature

Upon nuclear waste canister failure and contact of spent nuclear fuel with groundwater, the UO2 matrix of spent fuel will interact with oxidants in the groundwater generated by water radiolysis. Bicarbonate (HCO3-) is often found in groundwater, and the H2O2 induced oxidative dissolution of UO2 in bicarbonate solution has previously been studied under various conditions. Temperatures in the repository at the time of canister failure will differ depending on the location, yet the effect of temperature on oxidative dissolution is unknown. To investigate, the decomposition rate of H2O2 at the UO2 surface and dissolution of UVI in bicarbonate solution (0.1, 1, 10 and 50 mM) was analysed at various temperatures (10, 25, 45 and 60 °C). At [HCO3-] ≥ 1 mM, the concentration of dissolved UVI decreased with increasing temperature. This was attributed to the formation of UVI-bicarbonate species at the surface and a change in the mechanism of H2O2 decomposition from oxidative to catalytic. At 0.1 mM, no obvious correlation between temperature and U dissolution was observed, and thermodynamic calculations indicated this was due to a change in the surface species. A pathway to explain the observed dissolution behaviour of UO2 in bicarbonate solution as a function of temperature was proposed.

Mixed H2O/H2 plasma-induced redox reactions of thin uranium oxide films under UHV conditions

X-ray photoelectron spectroscopy (XPS) has been used to study the effect of mixed H2O/H2 gas plasma on the surfaces of UO2, U2O5 and UO3 thin films at 400 °C. The experiments were performed in situ under ultra-high vacuum conditions. Deconvolution of the U4f7/2 peaks into U(iv), U(v) and U(vi) components revealed the surface composition of the films after 10 min plasma exposure as a function of H2 concentration in the feed gas of the plasma. Some selected films (unexposed and exposed) were also analysed using grazing-incidence X-ray diffraction (GIXRD). The XPS results show that U(v) is formed as a major product upon 10 minutes exposure of UO3 by a mixed H2O/H2 plasma in a fairly wide H2 concentration range. When starting with U(v) (U2O5), rather high H2 concentrations are needed to reduce U(v) to U(iv) in 10 minutes. In the plasma induced oxidation of UO2, U(v) is never observed as a major product after 10 minutes and it would seem that once U(v) is formed in the oxidation of UO2 it is rapidly oxidized further to U(vi). The grazing incidence X-ray diffraction analysis shows that there is a considerable impact of the plasma and heating conditions on the crystal structure of the films in line with the change of the oxidation state. This structural difference is proposed to be the main kinetic barrier for plasma induced transfer between U(iv) and U(v) in both directions.

Reducing uncertainties affecting the assessment of the long-term corrosion behavior of spent nuclear fuel

Reducing the uncertainties associated with extrapolation to very long term of corrosion data obtainable from laboratory tests on a relatively young spent nuclear fuel is a formidable challenge. In a geologic repository, spent nuclear fuel may come in contact with water tens or hundreds of thousands of years after repository closure. The corrosion behavior will depend on the fuel properties and on the conditions characterizing the near field surrounding the spent fuel at the time of water contact. This paper summarizes the main conclusions drawn from multiyear experimental campaigns performed at JRC-ITU to study corrosion behavior and radionuclide release from spent light water reactor fuel. The radionuclide release from the central region of a fuel pellet is higher than that from the radial periphery, in spite of the higher burnup and the corresponding structural modifications occurring at the pellet rim during irradiation. Studies on the extent and time boundaries of the radiolytic enhancement of the spent fuel corrosion rate indicate that after tens or hundreds of thousands of years have elapsed, very small or no contribution to the enhanced corrosion rate has to be expected from α radiolysis. A beneficial effect inhibiting spent fuel corrosion due to the hydrogen overpressure generated in the near field by iron corrosion is confirmed. The results obtained so far point toward a benign picture describing spent fuel corrosion in a deep geologic repository. More work is ongoing to further reduce uncertainties and to obtain a full description of the expected corrosion behavior of spent fuel.

On Mo-Ru-Tc-Pd-Rh-Te alloy particles extracted from spent fuel and their leaching behavior under Ar and H2 atmospheres

In the present work, we extracted Mo-Tc-Ru-Pd-Rh-Te alloy particles from spent fuel by a non-oxidative selective leaching of the fuel matrix, characterized them and studied their leaching behavior under different redox conditions. After selecting the optimal experimental conditions by using a synthetic alloy, the radioactive alloy particles were extracted from spent fuel and characterized by using SEM (scanning electron microscopy)-WDS (wavelength dispersive spectra). Micrometer sized alloy particles display very similar compositions, on average (wt.%): 32.7% Mo, 40.5% Ru, 7% Tc, 4.4% Rh, 12.2% Pd and 3.8% Te. The alloy particles were leached in 10 mM NaCl and 2 mM NaHCO3 solutions, purged with two kinds of gas mixtures. It was observed that, under purging with Ar + 0.03% CO2, the leaching rates of 99Tc and 100Mo were similar, about 1.5 ppb/day. The results obtained under purging with 89.97% Ar + 0.03% CO2 + 10% H2 indicate that the 4d alloy particles are very effective catalysts for reduction caused by hydrogen: the concentrations of 99Tc and 100Mo dropped to levels below 1 ppb. No significant leaching of Ru, Rh, Pd and Te was observed during purging with both gas mixtures.

The reduction of U(VI) by near field hydrogen in the presence of UO2(s)

In several recent works a large impact of dissolved hydrogen on spent fuel dissolution has been observed: the solution concentrations of all redox sensitive components of the spent fuel decrease with time, while gas phase analyses indicate levels of radiolytic oxygen below the detection limit. This indicates that at the relatively low temperatures of these studies (≤70 °C) hydrogen is activated. The aim of this study is to test one of the proposed mechanisms for hydrogen activation, namely by UO2(s) surfaces, which may act as hydrogen catalysts. Dissolved U(VI) carbonate species were used as oxidized species in order to test the reducing ability of hydrogen in the presence and absence of UO2(s). In most cases parallel experiments with the same conditions, but under Ar(g) atmosphere were carried out as blanks. Quartz lined autoclaves avoiding any contact of the solution with metallic parts were used. The experiments were carried out in two stages: first the stability of the (10-7-10-5) M U(VI) in 10 mM NaCl, 2 mM NaHCO3 solutions in the presence of dissolved hydrogen was tested. Then a gold net basket containing UO2(s) fragments was introduced in the autoclave and the concentration of U(VI) was followed at different time intervals by ICP-MS. Various dissolved hydrogen concentrations, temperatures and U(VI) concentrations were used, always spanning the range of these parameters in a proposed deep rock repository. Especially at the lowest U(VI) concentrations tested (∼100 ppb) the activation of dissolved hydrogen by UO2(s) surfaces is clearly visible. As a conclusion of this work it may be stated that in the temperature range (20-70 °C) investigated a) dissolved hydrogen does not reduce U(VI) carbonate species in the absence of a catalyst and b) in the presence of UO2(s) surfaces dissolved hydrogen reduces U(VI) in carbonate solutions, very probably to UO2(s).

The geochemical behaviour of Tc, Np and Pu in spent nuclear fuel in an oxidizing environment

Spent fuel from commercial nuclear reactors consists mainly of uranium oxide. However, the changes that occur during reactor operations have a profound effect on chemical and physical properties of this material. Heat build-up in the fuel pellet during reactor operations can cause redistribution of fission products. The fission products may aggregate in one of three types of precipitates; gaseous, metallic, or oxide, depending on the burn-up and in-core treatment. Radiation damage and variations in fission and neutron capture yields across the fuel pellets lead to Pu enrichment and increased porosity with increasing burn-up. A more porous surface may make the fuel more susceptible to oxidative dissolution. As the level of actinides and fission products increases, the fuel may become more resistant to oxidation. These changes may limit the usefulness of natural uraninite (UO2) analogues for predicting the geological behaviour of spent fuel disposed in a high-level waste (HLW) repository. In this Chapter, an overview of spent fuel microstructure, radiolytic effects, and alteration processes is presented. Evidence for Np incorporation into U6+ phases, the nature of Pu surface precipitates on spent fuel, and evidence for the preferential removal of 4d-metals from ε-particles in corroded spent fuel is discussed. Understanding the potential mechanisms of radionuclide attenuation through sorption and/or incorporation requires techniques with both high spatial resolution and excellent elemental sensitivity.