2 edition of Dissolution of Irradiated uo2 Fuel Under Hydrothermal Oxidizinig Conditions. found in the catalog.
Dissolution of Irradiated uo2 Fuel Under Hydrothermal Oxidizinig Conditions.
Atomic Energy of Canada Limited.
|Series||Technical record (Atomic Energy of Canada Ltd) -- 128|
|Contributions||Johnson, L., Burns, K., Joling, H.|
Journal Article: Dissolution of Irradiated Commercial UO2 Fuels in Ammonium Carbonate and Hydrogen Peroxide. In this work we have studied oxidative dissolution of pure UO 2 and ADOPT (UO 2 doped with Al and Cr) pellets using H 2 O 2 and gammaradiolysis to induce the process. There is a small but significant difference in the oxidative dissolution rate of UO 2 and ADOPT pellets, respectively. However, the difference in oxidative dissolution yield is insignificant.
Journal Article: The effect of ion irradiation on the dissolution of UO 2 and UO 2-based simulant fuel. It is shown that both the studied reactive solutes (under oxygen free conditions) and the combination of Pd inclusions and H 2 inhibit the dissolution. Calculations (based on the fuel inventory) show that 1 µM Fe(II)(aq) decreases the dissolution rate by a factor of ~50 and that 1 ppm surface coverage of ε-particles is sufficient to.
Leaching experiments using spent nuclear fuel were also performed on the two types of fuel showing the same behavior as the unirradiated pellets, i.e., Oxidative dissolution of ADOPT compared to standard UO2 fuel H 2 O 2 induced dissolution of UO 2. Comparison of the Leaching Behaviour of Irradiated Fuel, SIMFUEL, and Non-Irradiated UO2 under Oxic Conditions. SERRANO J Articles in periodicals and books: JRC EPMA of Melted UO2 Fuel Rods Irradiated to a Burn-up of 23 GWd/tU. Influence of Low Temperature Air Oxidation on the Dissolution Behaviour of UO2 and MOX Spent Fuel.
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The results indicate that under the present experimental conditions, UO2 oxidation by H2O2 and UO2 dissolution are connected as a sequential reaction: for a given H2O2 concentration, UO2 oxidation.
The rate of dissolution of unirradiated UO2 fuel pellets under mildly oxidizing conditions has been investigated. A static leaching procedure was used Cited by: This proves that the UO2 surface constitutes a major redox buffer capacity to prevent radiolytic oxidation under repository conditions.
Introduction The dissolution of UO2, the fuel of light water reactors, has been investigated under non oxidizing [1,2] and oxidizing [3,4] conditions. The effect of dis- solved oxidants and radiolyticaUy Cited by: Dissolution experiments were conducted in single-pass flow-through (SPFT) mode under anoxic conditions (irradiation damage on the dissolution of the UO 2 matrix with data collection capturing six minute intervals for several by: 5.
Dissolution experiments were conducted in single-pass flow-through (SPFT) mode under anoxic conditions (irradiation damage on the dissolution. CHEMISTRY OF UO2 FUEL DISSOLUTION IN RELATION TO THE DISPOSAL OF USED NUCLEAR FUEL by S. Sunder and D.V. Shoesmith ABSTRACT This report reviews the chemistry of U02 dissolution under conditions relevant to the disposal of used nuclear fuel in a geological vault.
Also included are the UO 2 dissolution-rate equations as a function of pH, carbonate concentration, and oxidant concentration under oxidising conditions.
These experimental equations make it possible to model independent experiments performed with both non-irradiated and irradiated UO 2. The degradation behavior in high pressure water of UN and UN + (5–10 w%) UO 2 monolithic pellets fabricated from UN synthesized via a hydride-dehydride-nitride thermal process was investigated.
Sintered pellets (>90% theoretical density) were subjected to hydrothermal oxidation in a water-filled static autoclave at temperatures ranging from to °C and pressures to MPa. The effects of H2 and Y2O3 on the oxidative dissolution of UO2 under gamma irradiation are similar to those found in experiments with H2O2.
of irradiated fuel under different conditions: with. A kinetic model, which takes into account the mechanism of UO2 oxidation, is more appropriate to estimate dissolution rates of UO2 fuel for redox conditions more oxidizing than mV vs SCE.
We propose and test a disposition path for irradiated nuclear fuel using ammonium carbonate and hydrogen peroxide media. We demonstrate on a 13 g scale that >98% of the irradiated fuel dissolves.
Subsequent expulsion of carbonate from the dissolver solution precipitates >95% of the plutonium, americium, and curium and substantial amounts of fission products, effectively partitioning the fuel.
Mobilisation of uranium in geologic environments from UO 2 solid phases usually takes place by oxidative dissolution involving a change of U oxidation state from +IV to +VI; however, anoxic or reducing geochemical conditions are expected in many of the planned European disposal sites.
This work investigates potential alteration mechanisms of UO 2 in contact with groundwater ions (Ca 2+. Under slightly acidic pH conditions, the formation of an oxidized UO2+x phase was not observed on the surface and did not occur in the radiolysis dissolution mechanism of the fuel matrix.
containing spent fuel can be assessed. The oxidation of UO, leads to the formation of progressively higher oxidation states of uranium and an oxidized surface layer with a composition of U02+x (where 0 I x 11).
This phase is sparingly soluble under relatively reducing conditions, where the oxidation state remains below UO Articles in periodicals and books Abstract: Mobilisation of uranium in geologic environments from UO2 solid phases usually takes place by oxidative dissolution involving a change of U oxidation state from IV to VI; however, anoxic or reducing geochemical conditions are expected in many of the planned European disposal sites.
The oxidative dissolution of uranium(IV) dioxide powder at room temperature in aqueous carbonate media has been investigated. Kinetic studies evaluating the efficacy of various oxidants, including K2S2O8, NaOCl, and H2O2, for dissolving UO2 in alkaline solution have been performed, with H2O2 exhibiting the most rapid initial dissolution at M oxidant concentrations.
The dissolution of unirradiated UO2 fuel pellets under simulated disposal conditions. Nuclear and Chemical Waste Management5 (2), DOI: /X(84) J Bruno, R S Forsyth, L O Werme. Spent UO 2 -fuel Dissolution.
In this work, the efficiency of one- and two-electron oxidants in oxidative dissolution of UO 2 has been investigated. This was accomplished by measuring the U(VI)-concentration in solution after exposing UO 2-powder to controlled amounts of oxidants in aqueous oxidants used in this study are H 2 O 2, IrCl 6 2−, CO 3 • and OH •.H 2 O 2 acts as a two-electron oxidant while.
Oxidation and dissolution of UO2 in bicarbonate media: Implications for the spent nuclear fuel oxidative dissolution mechanism. Journal of Nuclear Materials(), DOI: /t The rate of uranium dioxide dissolution under oxidizing conditions in carbonate/bicarbonate media was found to be directly proportional to the total hydrogen carbonate concentration by Grandstaff.
Generally, within the pH range from to the rate of dissolution of uranium dioxide is independent of carbonate/bicarbonate ratio because the. The behavior of spent nuclear fuel under geol. conditions is a major issue underpinning the safety case for final disposal.
This work describes the prepn. and characterization of a non-radioactive UO2 fuel analog, CeO2, to be used to investigate nuclear fuel dissoln.
under realistic repository conditions as part of a developing EU research program.Dissolution and reaction of yttria‐stabilized zirconia (YSZ) single crystals were investigated in various solutions at ° to °C under MPa. YSZ crystals were not corroded in pure H 2 O and neutral solutions such a LiF, LiCl, NaNO 3, KCl, KBr; K 2 SO 4, and Na 2 SO 4 even under severe conditions at °C, MPa.Dissolution rates of different glasses in buffered solutions of constant pH of 7 were measured by weight change, profilometry, and ion implantation with Rutherford backscattering.
Corrosion behavior of silicon oxycarbide-based ceramic nanocomposites under hydrothermal conditions, International Journal of Materials Thomas Chassé.