Dissolution of irradiated fuel: a radiolytic mass balance study (original) (raw)
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The interaction of molecular hydrogen with α-radiolytic oxidants on a (U,Pu)O2 surface
Journal of Nuclear Materials, 2018
In order to assess the impact of a-radiolysis of water on the oxidative dissolution of spent fuel, an unirradiated, annealed MOX fuel pellet with high content of Pu (~24 wt%), and a specific a-activity of 4.96 GBq/g MOX , was leached in carbonate-containing solutions of low ionic strength. The high Pu content in the pellet stabilizes the (U,Pu)O 2 (s) matrix towards oxidative dissolution, whereas the a-decays emitted from the surface are expected to produce~3.6 Â 10 À7 mol H 2 O 2 /day, contributing to the oxidative dissolution of the pellet. Two sets of leaching tests were conducted under different redox conditions: Ar gas atmosphere and deuterium gas atmosphere. A relatively slow increase of the U and Pu concentrations was observed in the Ar case, with U concentrations increasing from 1$10-6 M after 1 h to~7 Â 10-5 M after 58 days. Leaching under an atmosphere starting at 1 MPa deuterium gas was undertaken in order to evaluate any effect of dissolved hydrogen on the radiolytic dissolution of the pellet, as well as to investigate any potential recombination of the a-radiolytic products with dissolved deuterium. For the latter purpose, isotopic analysis of the D/H content was carried out on solution samples taken during the leaching. Despite the continuous production of radiolytic oxidants, the concentrations of U and Pu remained quite constant at the level of~3 Â 10-8 M during the first 30 days, i.e. as long as the deuterium pressure remained higher than 0.8 MPa. These data rule out any oxidative dissolution of the pellet during the first month. The un-irradiated MOX fuel does not contain metallic ε-particles, hence it is mainly the interaction of radiolytic oxidants and dissolved deuterium with the surface of the mixed actinide oxide that causes the neutralization of the oxidants. This conclusion is supported by the steadily increasing levels of HDO measured in the leachate samples.
Journal of Nuclear Materials, 2005
Calibration and testing are inherent aspects of any modelling exercise and consequently they are key issues in developing a model for the oxidative dissolution of spent fuel. In the present work we present the outcome of the calibration process for the kinetic constants of a UO 2 oxidative dissolution mechanism developed for using in a radiolytic model. Experimental data obtained in dynamic leaching experiments of unirradiated UO 2 has been used for this purpose. The iterative calibration process has provided some insight into the detailed mechanism taking place in the alteration of UO 2 , particularly the role of Å OH radicals and their interaction with the carbonate system. The results show that, although more simulations are needed for testing in different experimental systems, the calibrated oxidative dissolution mechanism could be included in radiolytic models to gain confidence in the prediction of the long-term alteration rate of the spent fuel under repository conditions.
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 oxidative dissolution of unirradiated UO2 by hydrogen peroxide as a function of pH
Journal of Nuclear Materials, 2005
The dissolution of non-irradiated UO 2 was studied as a function of both pH and hydrogen peroxide concentration (simulating radiolytic generated product). At acidic pH and a relatively low hydrogen peroxide concentration (10 À5 mol dm À3 ), the UO 2 dissolution rate decreases linearly with pH while at alkaline pH the dissolution rate increases linearly with pH. At higher H 2 O 2 concentrations (10 À3 mol dm À3 ) the dissolution rates are lower than the ones at 10 À5 mol dm À3 H 2 O 2 , which has been attributed to the precipitation at these conditions of studtite (UO 4 AE 4H 2 O, which was identified by X-ray diffraction), together with the possibility of hydrogen peroxide decomposition. In the literature, spent fuel dissolution rates determined in the absence of carbonate fall in the H 2 O 2 concentration range 5 · 10 À7 -5 · 10 À5 mol dm À3 according to our results, which is in agreement with H 2 O 2 concentrations determined in spent fuel leaching experiments.
Release of Radiotoxic Elements from High Burn-Up UO2 and MOX Fuel in a Repository
MRS Proceedings, 2000
ABSTRACTIn a spent fuel repository the processes that govern the release of radionuclides are dissolution and transport in a possible groundwater flow. The cladding will be the last barrier before the water comes into contact with the fuel, namely with the outer rim of the pellet. Here the heterogeneity of the material due to the irradiation process is responsible for a complex release process. Fission products and minor actinides inventories are considerably higher at the pellet periphery as a result of increased epithermal neutron capture and of migration in the case of the volatile fission products.The present paper gives a review of experimental activities at the Institute for Transuranium Elements (ITU). Both single effects studies and integral tests are carried out to study the behavior of spent fuel under storage conditions.Leaching of irradiated UO2 (up to 50 GWd/tU) and MOX (up to 25 GWd/tU) fuel rods with preset cladding defects at 100°C under anoxic or reducing conditions...
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.
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:
Radiation Induced Spent Nuclear Fuel Dissolution under Deep Repository Conditions
Environmental Science & Technology, 2007
The dynamics of spent nuclear fuel dissolution in groundwater is an important part of the safety assessment of a deep geological repository for high level nuclear waste. In this paper we discuss the most important elementary processes and parameters involved in radiation induced oxidative dissolution of spent nuclear fuel. Based on these processes, we also present a new approach for simulation of spent nuclear fuel dissolution under deep repository conditions. This approach accounts for the effects of fuel age, burn up, noble metal nanoparticle contents, aqueous H 2 and HCO 3concentration, water chemistry, and combinations thereof. The results clearly indicate that solutes consuming H 2 O 2 and combined effects of noble metal nanoparticles and H 2 have significant impact on the rate of spent nuclear fuel dissolution. Using data from the two possible repository sites in Sweden, we have employed the new approach to estimate the maximum rate of spent nuclear fuel dissolution. This estimate indicates that H 2 produced from radiolysis of groundwater alone will be sufficient to inhibit the dissolution completely for spent nuclear fuel older than 100 years.
Water corrosion of spent nuclear fuel: radiolysis driven dissolution at the UO2/water interface
Faraday Discuss., 2015
X-ray diffraction has been used to probe the radiolytic corrosion of uranium dioxide. Single crystal thin films of UO2 were exposed to an intense X-ray beam at a synchrotron source in the presence of water, in order to simultaneously provide radiation fields required to split the water into highly oxidising radiolytic products, and to probe the crystal structure and composition of the UO2 layer, and the morphology of the UO2/water interface. By modeling the electron density, surface roughness and layer thickness, we have been able to reproduce the observed reflectivity and diffraction profiles and detect changes in oxide composition and rate of dissolution at the Ångström level, over a timescale of several minutes. A finite element calculation of the highly oxidising hydrogen peroxide product suggests that a more complex surface interaction than simple reaction with H2O2 is responsible for an enhancement in the corrosion rate directly at the interface of water and UO2, and this may ...
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.