[en] Alpha, beta, and gamma dose rates in water, in contact with the reference used fuel in the Canadian Nuclear Fuel Waste Management Program (i.e., Bruce A CANDU fuel, burnup 685 GJ/kg U), are calculated as a function of cooling time. Procedures to calculate the dose rates for used CANDU fuels of different burnups from results obtained for the reference fuel are described. These procedures can be adapted to estimate dose rates for fuels other than CANDU fuel. The dose rate information is needed to compare the results of leaching and corrosion experiments, conducted by different groups, with used fuels of different burnups and/or cooling times and to predict the effects of radiolysis of groundwater on used fuel oxidation and dissolution in a disposal vault. (author)

[en] Literature data on the effect of water radiolysis products on spent fuel oxidation and dissolution have been reviewed. Effects of γ-radiolysis, α-radiolysis and dissolved O₂ or H₂O₂ in unirradiated solutions have been discussed separately. Also the effect of carbonate in γ-irradiated solutions and radiolysis effects on leaching of spent fuels have been reviewed. In addition a radiolysis model for calculation of corrosion rates of UO₂, presented previously, has been discussed. The model has been shown to give a good agreement between calculated and measured corrosion rates in the case of γ-radiolysis and in unirradiated solutions of dissolved oxygen or hydrogen peroxide. The model has failed to predict the results of α-radiolysis. In a recent study it was shown that the model gave a good agreement with measured corrosion rates of spent fuel exposed in deionized water

[en] Alpha, beta and gamma dose rates in water, in contact with the reference used fuel in the Canadian Nuclear Fuel Waste Management Program, i.e., Bruce A CANDU fuel, burnup 685 GJ/kg U, are calculated as a function of cooling time. Procedures are described to calculate the dose rates for used CANDU fuels of different burnups from results obtained for the reference fuel. This information is needed to compare the results of leaching and corrosion experiments carried out with used CANDU fuels of different burnups and/or cooling times and to predict the effects of radiolysis of groundwater on used fuel oxidation and dissolution in a disposal vault. (author). 37 refs., 4 tabs., 7 figs

[en] Corrosion of UN in water was investigated as a function of pH and temperature using X-ray photoelectron spectroscopy (XPS), X-ray diffraction (XRD), and by measuring the amount of ammonia formed due to its corrosion. The XPS results indicate that a freshly fractured surface of UN is quickly converted to U0₂ on exposure to liquid water or water vapours at ambient temperatures. These results show that UN is unstable in contact with water. The corrosion rate of UN is estimated to be ≥40 μmol·m^-2·h^-1 in deaerated water at ∼92^oC. There was no significant difference in corrosion rates measured in water at initial pHs of ∼6 and ∼10.3. These results contradict the literature reports stating that UN is stable in contact with boiling water. The implications of these results on the suitability of UN as a nuclear fuel for reactors are discussed. (author)

[en] Dissolution and corrosion of (Th,U)O₂ fuel was investigated at room temperature and near 100 degrees C in near neutral and acidic water (pH 3) to evaluate the suitability of irradiated UO₂-doped thoria as a waste form for direct geological disposal. X-ray photoelectron spectroscopy and X-ray diffraction were used to study oxidation of (Th,U)O₂ fuel. The uranium in the surface of (Th,U)O₂ fuel undergoes oxidation similar to that observed in UO₂ fuel under similar conditions. Nevertheless, the dissolution rate of uranium from (Th,U)O₂ fuel in aerated solutions is much lower than that from UO₂ fuel under similar conditions. (author)

[en] UO₂ disks were heated at 150 deg C in air, in O₂ with 60% saturated steam, and in Ar with 60% saturated steam atmospheres for ∼2 years in gamma fields equivalent to those associated with 10- to 20-year-old used CANDU fuel. Surface analysis of the disks, using X-ray photoelectron spectroscopy, X-ray diffraction and scanning electron microscopy, shows formation of U₃O₈ on the UO₂ disks exposed to air or O₂. This is the first report of the formation of U₃O₈ on UO₂ by air oxidation at such a low temperature. The rate of U₃O₈ formation by dry air oxidation of UO₂ at 150 deg C and in gamma fields of dose rate ∼15 Gy·h^-1 is very low. The presence of water vapour along with O₂ increases the oxidation of UO₂ in gamma fields, leading to the formation of U⁶⁺ phases, e.g., UO₃·xH₂O along with U₃O₈. On the other hand, UO₂ disks did not suffer any oxidation by water vapour radiolysis at 150 deg C in an O₂-free (60% saturated steam in Ar) atmosphere. These are important observations for the dry storage of used fuel because the oxidation of UO₂ to U₃O₈ is accompanied by a volume expansion, which could result in splitting of the Zircaloy cladding and powdering of the fuel matrix. (author). 46 refs., 7 tabs., 16 figs

No abstract available

[en] A code to calculate thermodynamic functions of an ideal gas (at 1 atm pressure) from its spectroscopic data using statistical mechanics methods was commissioned, tested and verified. The code was used to calculate the thermodynamic functions for three fission-product oxides, Ce0₂, Ce₂0₂ and NdO₂, for temperatures between 100 K and 3000 K and 0.1 MPa. High-temperature thermodynamic data of fission products are needed to calculate the source terms under accident conditions using fuel safety codes such as CHMWRK and F*A*C*T. (author)

[en] This report reviews the chemistry of UO₂ dissolution under conditions relevant to the disposal of used nuclear fuel in a geological vault. It provides the chemical understanding necessary for selecting the most appropriate model for estimating UO₂ fuel dissolution rates in a nuclear waste disposal vault. The report briefly describes the solid-state structures of various uranium oxides; discusses the nature and mechanism of UO₂ oxidation and dissolution in groundwaters; summarizes the factors affecting UO₂ dissolution under oxidizing conditions; discusses the impact of various oxidants and water radiolysis on UO₂ oxidation and dissolution; briefly comments on the effects of vault chemistry and secondary phase formation on the dissolution process; discusses the physical properties of UO₂ that may influence the kinetics of dissolution; and describes our approach for developing a kinetic model of UO₂ dissolution under oxidizing conditions

[en] A model to predict the dissolution of UO₂ fuel under both oxidizing and non-oxidizing conditions is presented and compared with other available models for fuel dissolution. Dissolution rates under oxidizing conditions are predicted by extrapolating steady-state electrochemical currents for the anodic dissolution of UO₂ to the corrosion potentials measured in solutions containing various oxidants, including dissolved oxygen, hydrogen peroxide, and the products of the gamma or alpha radiolysis of water. For non-oxidizing conditions, the dissolution rate of UO₂ is not well known. Attempts to measure this rate are fraught with difficulties, and the published values are difficult to rationalize within the framework of our model. Consequently, we briefly reviewed the literature on the dissolution of similar p-type semiconducting oxides and chose to estimate the chemical dissolution rate of UO₂ by analogy to the well-studied oxide NiO. In this manner we have managed to establish a threshold rate below which the rate of oxidative dissolution becomes negligible in comparison with the rate of chemical dissolution. This threshold agrees quite well with that established electrochemically. Using these extrapolated rates we predict that the rate for oxidative dissolution of CANDU (CANada Deuterium Uranium) fuel due to gamma radiolysis will fall below this threshold after ∼ 200 a, a time period that is short in comparison with the anticipated lifetimes of titanium waste containers, which are expected to last for a period greater than ∼ 1200 a. For dissolution due to alpha radiolysis, oxidative rates are uncertain, but could be above this threshold for a period of 500 to 10 000 a for CANDU fuel, and 500 to 30 000 a for pressurized water reactor (PWR) fuel. The uncertainty in these ranges reflects the poor quality and limited number of corrosion potential measurements in the presence of alpha radiolysis