Dissolution kinetics of Indian PHWR natural UO2 fuel pellets in nitric acid – Effect of initial acidity and temperature (original) (raw)
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Dissolution of intact UO2 pellet in batch and rotary dissolver conditions
Journal of Radioanalytical and Nuclear Chemistry, 2014
Comparative dissolution of intact un-irradiated UO 2 pellet of PHWR fuel dimensions was performed in batch and dynamic rotary dissolver conditions in aqueous nitric acid solutions at elevated temperatures. The extent of dissolution was estimated by determining the uranium concentration of the resulting aqueous solution. It was observed that rate of dissolution was much faster in dynamic conditions as compared to static batch conditions.
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.
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.
Journal of Nuclear Materials, 2005
To evaluate the release of uranium from natural ore deposits, spent nuclear fuel repositories, and REDOX permeable reactive barriers (PRB), knowledge of the fundamental reaction kinetics associated with the dissolution of uranium dioxide is necessary. Dissolution of crystalline uranium (IV) dioxide under environmental conditions has been studied for four decades but a cardinal gap in the published literature is the effect of pH and solution saturation state on UO 2 (cr) dissolution. To resolve inconsistencies, UO 2 dissolution experiments have been conducted under oxic conditions using the single-pass flow-through system. Experiments were conducted as a function of total dissolved carbonate ð½CO À3 3 T Þ from 0.001 to 0.1 M; pH from 7.5 to 11.1; ratio of flow-through rate (q) to specific surface area (S), constant ionic strength (I) = 0.1 M, and temperatures (T) from 23 to 60°C utilizing both powder and monolithic specimens. The results show that UO 2 dissolution varies as a function of the ratio q/S and temperature. At values of log 10 q/S > À7.0, UO 2 dissolution becomes invariant with respect to q/S, which can be interpreted as evidence for dissolution at the forward rate of reaction. The data collected in these experiments show the rate of UO 2 dissolution increased by an order of magnitude with a 30°C increase in temperature. The results also show the overall dissolution rate increases with an increase in pH and decreases as the dissolved uranium concentration approaches saturation with respect to secondary reaction products. Thus, as the value of the reaction quotient, Q, approaches equilibrium, K, (with respect to a potential secondary phase) the dissolution rate decreases. This decrease in dissolution rate (r) was also observed when comparing measured UO 2 dissolution rates from static tests where r = 1.7 ± 0.14 · 10 À8 mol m À2 s À1 to the rate for flow-through reactors where r = 3.1 ± 1.2 · 10 À7 mol m À2 s À1 . Thus, using traditional static test methods can result in an underestimation of the true forward rate of UO 2 (cr) dissolution. These results illustrate the importance of pH, solution saturation state, and the concentration of dissolved carbonate on the release of uranium from UO 2 in the natural environment. Published by Elsevier B.V.
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.