The release of radioactive gases from uranium dioxide during irradiation

The steady-state release of radioactive rare gases has been measured during the isothermal irradiation of small-grained, largegrained and single-crystal UO₂. The experiment showed that (a) gases with half-lives less than five days exhibit no indication of enhanced grain-boundary diffusion; (b) the rate of release is a complex function of rating, but within the range of the experiment, the release rate was independent of rating; (c) the release kinetics for each gas are unique with no common relationship between fractional release and the gas decay constant; (d) the results show that the diffusion coefficient of xenon is ten times that for krypton at a temperature of 1450°C.     

The inhibiting effects of hydrogen on the corrosion of uranium dioxide under nuclear waste disposal conditions

A number of electrochemical experiments were employed to investigate the effects of hydrogen on the corrosion of UO₂ under nuclear waste disposal conditions. A combination of corrosion potential (E_CORR) measurements and cyclic voltammetry have indicated that dissolved hydrogen can polarize the UO₂ surface to reducing potentials; i.e., to E_CORR values more negative then those observed under anoxic (argon-purged) conditions. A comparison of the behaviours of SIMFUEL specimens with and without incorporated noble metal ε-particles indicates that these particles may act as catalytic electrodes for H₂ oxidation, H₂ ↔ 2e⁻ + 2H⁺. It is the galvanic coupling of these particles to the UO₂ matrix which suppresses the fuel corrosion potential.     

The characteristics of fission gas release from monocrystalline uranium dioxide during irradiation

Release rates for ^85mKr, ⁸⁷Kr, ⁸⁸Kr, ¹³³Xe, ¹³⁵xe and ¹³⁸Xe were measured in the temperature range 700–1550°C. The data were analysed in terms of diffusion of the rare gases and their halogen precursors. The diffusion coefficients for xenon and iodine were found to be similar whilst krypton also had a similar mobility at ~1200°C but otherwise diffused more slowly. Bromine had a high mobility compared with the rare gases (X 200).     

Secondary hydriding criteria under irradiation conditions

Reliable criteria of secondary hydriding failures are important to assure safe operation of nuclear fuel in LWR power units. The present paper reviews available data on massive hydriding of zirconium-based claddings covering out-of-pile studies and in-pile tests in research reactors. Analyses of these experimental data give evidence that threshold conditions leading to the onset of massive hydriding are drastically changed under irradiation. The changes are caused mainly by irradiation damage of oxygen sublattice in ZrO₂ by fission fragments leaving the periphery of fuel pellets. The tests in research reactors provide a basis to develop a parametric dependency which relates the threshold of massive hydriding to composition of steam-hydrogen mixture, irradiation dose rate and temperature.     

Observations on the low-temperature creep properties of uranium carbide under irradiation

The creep behaviour of 97% dense hyperstoichiometric UC has been examined during irradiation in three-point bend tests carried out at 450°C up to a dose of $\text{1.65 × 10}{}^{26}\text{fissions/m}{}^3$. A rapid decrease in measured strain rate with dose was observed at each stress level, nominally steady-state creep being established above $\text{≈ 1 × 10}{}^{26}\text{fissions/m}{}^3$ when the creep rate was a factor of 8 lower than that observed in UO₂ irradiated under identical conditions. Creep rates were found to be directly proportional to stress at high doses. Comparison of results from this test with data from other experiments up to $\text{2 × 10}{}^{25}\text{fissions/m}{}^3$ in compression and tension indicates little variation in the radiation-creep constant between 450°C and 800°C. The creep rate for UC, much lower than that observed in UO₂, is consistent with recently reported determinations of the effective uranium self-diffusion coefficients under irradiation in those materials.     

Coefficient of static friction of the uranium dioxide-zirconium alloy pair under irradiation

Moscow Engineering Physics Institute. Translated from Atomnaya Énergiya, Vol. 71, No. 5, pp. 463–466, November, 1991.     

Equation of state of uranium dioxide

An equation for the free energy of formation of $\text{UO}{}_2\text{(l)}$ is derived. Using this equation, a new calculational method for estimating consistent values for the vapor pressure of a liquid compound has been applied to uranium dioxide. Equations for the partial pressure of all vapor species are calculated. Values of the liquid density to the critical point are presented.     

The influence of precipitation conditions on the properties of ammonium diuranate and uranium dioxide powders

Continuous precipitation in one and two stages was investigated to determine the effect of precipi nation variables on the properties of ADU precipitates and the sinterability of subsequent UO₂ powders in pellet fabrication. The pH at which precipitation occurred was the most important parameter in determining the size of ADU agglomerates and the settling rate and filterability of the slurry. In two-stage precipitation, the ADU properties were determined by the proportion of uranium precipitated at different pH values. Washing nitrate from ADU appeared not to affect the properties of the subsequent UO₂ pswder but a significant decrease in filterability occurred when the ADU was extensively washed. When ADU was reduced to UO₂ at 600 °C the agglomerate structure of the ADU was retained. The larger the agglomerates in the UO₂ powder (lower pH of precipitation of precursor ADU) the less sinterable it was; pseudomorphs of large agglomerates were still discernible in the sintered pellets. Thus settling rate of the ADU slurry gave an early indication of the likely sinterability of the resultant UO₂ powder since both were functions of agglomerate size.     

The superplastic creep of uranium dioxide

In common with many other ceramics, uranium dioxide exhibits low creep ductility in both bending and compression loading modes. It is demonstrated that if the material and test conditions are suitably optimised, failure strains in compression of over 100% true strain are possible. The growth and reorientation of grain-boundary pores during such superplastic deformation are studied and a model to explain their behaviour is proposed.