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| 3. |
- Bauhn, Lovisa, 1981, et al.
(författare)
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The fate of hydroxyl radicals produced during H2O2 decomposition on a SIMFUEL surface in the presence of dissolved hydrogen
- 2018
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 507, s. 38-43
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Tidskriftsartikel (refereegranskat)abstract
- Over geologic timescales hydrogen peroxide will be one of the most important radiolytic oxidants challenging the spent fuel integrity in a deep repository. Consequently, the reaction between hydrogen peroxide and different kinds of UO 2 based materials has been the subject of several studies over recent decades. Parts of these studies have investigated the effect of dissolved hydrogen on this reaction, as large amounts of hydrogen are expected to be produced by anoxic corrosion of iron in the deep repositories. In some of the studies hydrogen has been shown to offset the radiolysis-driven oxidative dissolution of the fuel despite the expected inertia of hydrogen at repository temperatures. However, the underlying mechanism is primarily based on the effect of the metallic particles contained in the spent fuel. One clue to the mechanistic understanding is whether or not a reaction takes place between dissolved hydrogen and hydroxyl radicals adsorbed to a fuel surface resulting from the decomposition of H 2 O 2 . In the study presented here this reaction could be confirmed in an autoclave system with SIMFUEL, a hydrogen peroxide spiked solution, and deuterium gas. The results show that the studied reaction does not only occur, but accounts for a substantial part of the hydrogen peroxide consumption in the system. Only a very minor part, 0.02%, of the total consumed hydrogen peroxide caused oxidative dissolution of the SIMFUEL. The conclusion is supported by quantitative measurements of HDO, dissolved U in solution and O 2 in the gas phase.
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| 4. |
- Bauhn, Lovisa, 1981, et al.
(författare)
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The interaction of molecular hydrogen with α-radiolytic oxidants on a (U,Pu)O2 surface
- 2018
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 505, s. 54-61
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Tidskriftsartikel (refereegranskat)abstract
- In order to assess the impact of α-radiolysis of water on the oxidative dissolution of spent fuel, an un-irradiated, annealed MOX fuel pellet with high content of Pu (∼24 wt%), and a specific α-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 α-decays emitted from the surface are expected to produce ∼3.6 × 10 −7 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 × 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 α-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 × 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 ra diolytic 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.
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| 5. |
- Carbol, P., et al.
(författare)
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Corrosion of irradiated MOX fuel in presence of dissolved H-2
- 2009
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 392:1, s. 45-54
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Tidskriftsartikel (refereegranskat)abstract
- The corrosion behaviour of irradiated MOX fuel (47 GWd/tHM) has been studied in an autoclave experiment simulating repository conditions. Fuel fragments were corroded at room temperature in a 10 mM NaCl/2 mM NaHCO3 solution in presence of dissolved H-2 for 2100 days. The results show that dissolved H-2 in concentration 1 mM and higher inhibits oxidation and dissolution of the fragments. Stable U and Pu concentrations were measured at 7 x 10(-10) and 5 x 10(-11) M, respectively. Caesium was only released during the first two years of the experiment. The results indicate that the UO2 matrix of a spent MOX fuel is the main contributor to the measured dissolution, while the corrosion of the high burn-up Pu-rich islands appears negligible. (C) 2009 Elsevier B.V. All rights reserved.
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| 6. |
- Carbol, P., et al.
(författare)
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Hydrogen suppresses UO2 corrosion
- 2009
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Ingår i: Geochimica et Cosmochimica Acta. - : Elsevier BV. - 0016-7037 .- 0046-564X. ; 73:15, s. 4366-4375
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Tidskriftsartikel (refereegranskat)abstract
- Release of long-lived radionuclides such as plutonium and caesium from spent nuclear fuel in deep geological repositories will depend mainly on the dissolution rate of the UO2 fuel matrix. This dissolution rate will, in turn, depend on the redox conditions at the fuel surface. Under oxidative conditions UO2 will be oxidised to the 1000 times more soluble UO2.67. This may occur in a repository as the reducing deep groundwater becomes locally oxidative at the fuel surface under the effect of alpha-radiolysis, the process by which alpha-particles emitted from the fuel split water molecules. On the other hand, the groundwater corrodes canister iron generating large amounts of hydrogen. The role of molecular hydrogen as reductant in a deep bedrock repository is questioned. Here we show evidence of a surface-catalysed reaction, taking place in the H-2-UO2-H2O system where molecular hydrogen is able to reduce oxidants originating from alpha-radiolysis. In our experiment the UO2 surface remained stoichiometric proving that the expected oxidation of UO2.00 to UO2.67 due to radiolytic oxidants was absent. As a consequence, the dissolution of UO2 stopped when equilibrium was reached between the solid phase and U4+ species in the aqueous phase. The steady-state concentration of uranium in solution was determined to be 9 x 10(-12) M, about 30 times lower than previously reported for reducing conditions. Our findings show that fuel dissolution is suppressed by H-2. Consequently, radiotoxic nuclides in spent nuclear fuel will remain immobilised in the UO2 matrix. A mechanism for the surface-catalysed reaction between molecular hydrogen and radiolytic oxidants is proposed. (C) 2009 Elsevier Ltd. All rights reserved.
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| 7. |
- Cen, Peng, et al.
(författare)
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Metal extraction from a deep eutectic solvent, an insight into activities
- 2020
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Ingår i: Physical Chemistry Chemical Physics. - : Royal Society of Chemistry (RSC). - 1463-9084 .- 1463-9076. ; 22:19, s. 11012-11024
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Tidskriftsartikel (refereegranskat)abstract
- The solvent extraction of gold, palladium, technetium, indium and rhenium from both mixtures of a deep eutectic solvent with aqueous salt solutions and of two different aqueous salt solutions has been performed. Initially using gold and then the other metals the solvent extraction results were interpreted using an activity coefficient/function equation (specific ion interaction theory). The most important presented result is the addition of a new term to the specific ion interaction theory equation. Using the new model it will be possible to make predictions of the behaviour of a solvent extraction system where one of the liquid phases is a mixture of aqueous salts and the deep eutectic solvent.
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| 8. |
- Cui, Daqing, et al.
(författare)
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Surface mediated processes in the interaction of spent fuel α-doped UO2 with H2
- 2008
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Ingår i: Materials Research Society Symposium Proceedings. - 0272-9172. - 9781605110745 ; 1104, s. 87-99
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Konferensbidrag (refereegranskat)abstract
- 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 233U or 238Pu 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 ?-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 redox-sensitive radionuclides, such as Tc and the minor actinides. The concentrations of U are systematically lower than the values measured during UO2(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 UO2(s) pellets doped with 233U 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 UO2 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 ?- radiolysis.
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| 9. |
- Ekeroth, Ella, et al.
(författare)
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The effect of temperature and fuel surface area on spent nuclear fuel dissolution kinetics under H2 atmosphere
- 2020
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 531
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Tidskriftsartikel (refereegranskat)abstract
- In this work we present the results of two spent nuclear fuel leaching experiments in simulated granitic groundwater, saturated with hydrogen under various pressures. The results show a large impact of the dissolved hydrogen already at 1 bar H2 and room temperature on the release of both the uranium and of the fission products contained in the fuel matrix. Based on the results of this study and on published data with fuel from the same rod, the importance of the oxidative dissolution of spent fuel under repository conditions as compared to its non-oxidative dissolution is discussed. The XPS-spectra of the fuel surface before the tests and after long-term leaching under hydrogen are reported and compared to reduced UO2 and SIMFUEL surfaces. The overall conclusion is that in spite of the unavoidable air contamination, hydrogen pressures of 1 bar or higher counteract successfully the oxidative dissolution of the spent nuclear fuel. The stability of the 4d-element metallic particles during fuel leaching under such conditions is also discussed, based on data for their dissolution. The metallic particles are also stable under such conditions and are not expected to release their component metals during long-term fuel leaching. © 2020 The Authors
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| 10. |
- Fors, Patrik, 1979, et al.
(författare)
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Corrosion of high burn-up structured UO2 fuel in presence of dissolved H-2
- 2009
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 394:1, s. 1-8
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Tidskriftsartikel (refereegranskat)abstract
- The influence of high burn-up structured material on UO2 corrosion has been studied in an autoclave experiment. The experiment was conducted on spent fuel fragments with an average burn-up of 67 GWd/tHM. They were corroded in a simplified groundwater containing 33 mM dissolved H-2 for 502 days. All redox sensitive elements were reduced. The reduction continued until a steady-state concentration was reached in the leachate for U at 1.5 x 10(-10) M and for Pu at 7 x 10(-11) M. The instant release of Cs during the first 7 days was determined to 3.4% of the total inventory. However, the Cs release stopped after release of 3.5%. It was shown that the high burn-up structure did not enhance fuel corrosion. (C) 2009 Elsevier B.V. All rights reserved.
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