The effects of alpha-radiolysis on UO2 dissolution determined from batch experiments with 238Pu-doped UO2 (original) (raw)
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The effect of ion irradiation on the dissolution of UO2 and UO2-based simulant fuel
Journal of Alloys and Compounds, 2018
The aim of this work was to study the separate effect of fission fragment damage on the dissolution of simulant UK advanced gas-cooled reactor nuclear fuel in water. Plain UO 2 and UO 2 samples, doped with inactive fission products to simulate 43 GWd/tU of burn-up, were fabricated. A set of these samples were then irradiated with 92 MeV 129 Xe 23þ ions to a fluence of 4.8 Â 10 15 ions/cm 2 to simulate the fission damage that occurs within nuclear fuels. The primary effect of the irradiation on the UO 2 samples, observed by scanning electron microscopy, was to induce a smoothening of the surface features and formation of hollow blisters, which was attributed to multiple overlap of ion tracks. Dissolution experiments were conducted in single-pass flow-through (SPFT) mode under anoxic conditions (<0.1 O 2 ppm in Ar) to study the effect of the induced irradiation damage on the dissolution of the UO 2 matrix with data collection capturing six minute intervals for several hours. These time-resolved data showed that the irradiated samples showed a higher initial release of uranium than unirradiated samples, but that the uranium concentrations converged towards~10 À9 mol/l after a few hours. Apart from the initial spike in uranium concentration, attributed to irradiation induced surficial micro-structural changes, no noticeable difference in uranium chemistry as measured by X-ray electron spectroscopy or 'effective solubility' was observed between the irradiated, doped and undoped samples in this work. Some secondary phase formation was observed on the surface of UO 2 samples after the dissolution experiment.
Effect of external gamma irradiation on dissolution of the spent UO2 fuel matrix
Journal of Nuclear Materials, 2005
Leaching experiments were performed on UO 2 pellets doped with alpha-emitters (238/239 Pu) and on spent fuel, in the presence of an external gamma irradiation source (A 60 Co = 260 Ci, _ Dc ¼ 650 Gy h À1). The effects of a, b, c radiation, the fuel chemistry and the nature of the cover gas (aerated or Ar + 4%H 2) on water radiolysis and on oxidizing dissolution of the UO 2 matrix are quantified and discussed. For the doped UO 2 pellets, the nature of the cover gas clearly has a major role in the effect of gamma radiolysis. The uranium dissolution rate in an aerated medium is 83 mg m À2 d À1 compared with only 6 mg m À2 d À1 in Ar + 4%H 2. The rate drop is accompanied by a reduction of about four orders of magnitude in the hydrogen peroxide concentrations in the homogeneous solution. The uranium dissolution rates also underestimate the matrix alteration rate because of major precipitation phenomena at the UO 2 pellet surface. The presence of studtite in particular was demonstrated in aerated media; this is consistent with the measured H 2 O 2 concentrations (1.2 • 10 À4 mol L À1). For spent fuel, the presence of fission products (Cs and Sr), matrix alteration tracers, allowed us to determine the alteration rates under external gamma irradiation. The fission product release rates were higher by a factor of 5-10 than those of the actinides (80-90% of the actinides precipitated on the surface of the fragments) and also depended to a large extent on the nature of the cover gas. No significant effect of the fuel chemistry compared with UO 2 was observed on uranium dissolution and H 2 O 2 production in the presence of the 60 Co source in aerated conditions. Conversely, in Ar + 4%H 2 the fuel self-irradiation field cannot be disregarded since the H 2 O 2 concentrations drop by only three orders of magnitude compared with UO 2 .
The effect of fuel chemistry on UO2 dissolution
Journal of Nuclear Materials, 2016
The dissolution rate of both unirradiated UO 2 and used nuclear fuel has been studied by numerous countries as part of the performance assessment of proposed geologic repositories. In the scenario of waste package failure and groundwater contact with the fuel, the effects of variables such as temperature, dissolved oxygen, and water and fuel chemistry on the dissolution rates of the fuel are necessary to provide a quantitative estimate of the potential release over geologic time frames. The primary objective of this research was to determine the influence these parameters, with primary focus on the fuel chemistry, have on the dissolution rate of unirradiated UO 2 under oxidizing repository conditions and compare them to the rates predicted by current dissolution models.
Surface Mediated Processes in the Interaction of Spent Fuel or {alpha}-doped UO{sub 2} with H{sub 2}
2008
In most deep disposal concepts, large amounts of hydrogen are expected to be produced by the anoxic corrosion of massive iron containers. At repository temperatures, hydrogen is quite inert and is not expected to contribute to the redox capacity of the deep groundwaters. In several recent works, a large impact of dissolved hydrogen on the dissolution of the LWR or MOX fuel and UO 2 (s) doped with 233 U or 238 Pu has been observed. For hydrogen concentrations above a certain limit, the dissolution rates of these highly radioactive materials drop to very low values. A discussion of the results obtained with spent fuel or α-doped UO 2 in the presence of a range of hydrogen concentrations is presented. Typical for all measurements under such conditions are the very low long term concentrations of uranium and other redoxsensitive radionuclides, such as Tc and the minor actinides. The concentrations of U are systematically lower than the values measured during UO 2 (s) solubility measurements carried out in the presence of strong reducing agents. Measurements of the radiolytic oxygen after long leaching periods result in values below detection limit. The investigation of the surface of spent fuel or UO 2 (s) pellets doped with 233 U by XPS after long periods of testing shows absence of oxidation. The kinetics of the release of non-redox sensitive elements such as Sr and Cs, used to estimate fuel matrix dissolution rates, is also discussed. An attempt is made to propose potential mechanisms responsible for the observed behaviour, based mainly on data from studies on the interaction of water adsorbed on the surfaces of metal oxides or actinide oxides with radiation. Another important effect observed in recent studies is the existence of a threshold for the specific alpha activity below which no measurable influence of the alpha radiolysis on the uranium release from UO 2 is observed. The importance of such a threshold for the behaviour of spent fuel under repository conditions encompassing very long time scales will be discussed, as well as the necessity to better investigate the mechanisms of recombination reactions in a thin water layer on the surface of actinide oxides affected by α-radiolysis.
Surface Mediated Processes in the Interaction of Spent Fuel or alpha-doped UO2with H2
MRS Proceedings, 2008
In most deep disposal concepts, large amounts of hydrogen are expected to be produced by the anoxic corrosion of massive iron containers. At repository temperatures, hydrogen is quite inert and is not expected to contribute to the redox capacity of the deep groundwaters. In several recent works, a large impact of dissolved hydrogen on the dissolution of the LWR or MOX fuel and UO 2 (s) doped with 233 U or 238 Pu has been observed. For hydrogen concentrations above a certain limit, the dissolution rates of these highly radioactive materials drop to very low values. A discussion of the results obtained with spent fuel or α-doped UO 2 in the presence of a range of hydrogen concentrations is presented. Typical for all measurements under such conditions are the very low long term concentrations of uranium and other redoxsensitive radionuclides, such as Tc and the minor actinides. The concentrations of U are systematically lower than the values measured during UO 2 (s) solubility measurements carried out in the presence of strong reducing agents. Measurements of the radiolytic oxygen after long leaching periods result in values below detection limit. The investigation of the surface of spent fuel or UO 2 (s) pellets doped with 233 U by XPS after long periods of testing shows absence of oxidation. The kinetics of the release of non-redox sensitive elements such as Sr and Cs, used to estimate fuel matrix dissolution rates, is also discussed. An attempt is made to propose potential mechanisms responsible for the observed behaviour, based mainly on data from studies on the interaction of water adsorbed on the surfaces of metal oxides or actinide oxides with radiation. Another important effect observed in recent studies is the existence of a threshold for the specific alpha activity below which no measurable influence of the alpha radiolysis on the uranium release from UO 2 is observed. The importance of such a threshold for the behaviour of spent fuel under repository conditions encompassing very long time scales will be discussed, as well as the necessity to better investigate the mechanisms of recombination reactions in a thin water layer on the surface of actinide oxides affected by α-radiolysis.
MRS Proceedings, 2006
ABSTRACTSince 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 “lon...
Physical and Chemical Aspects of Radiation Induced Oxidative Dissolution of UO2
2006
The general subject of this thesis is oxidative dissolution of UO 2. The dissolution of UO 2 is mainly investigated because of the importance of the UO 2 matrix of spent nuclear fuel as a barrier against radionuclide release in a future deep repository. U(IV) is extremely insoluble under the reducing conditions prevalent in a deep repository, whereas U(VI) is more soluble. Hence, oxidation of the UO 2-matrix will affect its solubility and thereby its function as a barrier. In this thesis the relative efficiency of one-and two electron oxidants in dissolving UO 2 is studied. The oxidative dissolution yield of UO 2 was found to differ between one-and two-electron oxidants. At low oxidant concentrations the dissolution yields for one-electron oxidants are significantly lower than for two-electron oxidants. However, the dissolution yield for one-electron oxidants increases with increasing oxidant concentration, which could be rationalized by the increased probability for two consecutive one-electron oxidations at the same site and the increased possibility for disproportionation. This licentiate thesis is based on the following publications:
Journal of Nuclear Materials, 1996
The effect of H20 2, NaC10 and Fe on the dissolution of unirradiated UO2(s) in NaC1 5 mol kg-J has been studied at neutral to alkaline pH. Dissolution rates have been determined as a function of oxidant concentration. A general equation to correlate both parameters has been obtained: log r = (-8.0 + 0.2)+ log[Ox] 0"93+ 0.07. The values obtained have been compared to those given for spent fuel under the same experimental conditions. The effect of iron is similar in both unirradiated UO 2 and spent fuel with a final uranium concentration around 5 × 10 -8 mol kg-~ which corresponds to the solubility value of UO2(f) under reducing conditions.
MRS Advances, 2021
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.