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The general subject of this thesis is oxidative dissolution of UO 2. The dissolution of UO 2 is m... more The general subject of this thesis is oxidative dissolution of UO 2. The dissolution of UO 2 is mainly investigated because of the importance of the UO 2 matrix of spent nuclear fuel as a barrier against radionuclide release in a future deep repository. U(IV) is extremely insoluble under the reducing conditions prevalent in a deep repository, whereas U(VI) is more soluble. Hence, oxidation of the UO 2-matrix will affect its solubility and thereby its function as a barrier. In this thesis the relative efficiency of one-and two electron oxidants in dissolving UO 2 is studied. The oxidative dissolution yield of UO 2 was found to differ between one-and two-electron oxidants. At low oxidant concentrations the dissolution yields for one-electron oxidants are significantly lower than for two-electron oxidants. However, the dissolution yield for one-electron oxidants increases with increasing oxidant concentration, which could be rationalized by the increased probability for two consecutive one-electron oxidations at the same site and the increased possibility for disproportionation. This licentiate thesis is based on the following publications:
Journal of Nuclear Materials, 2019
The release of radionuclides from spent nuclear fuel upon contact with water is a central issue f... more The release of radionuclides from spent nuclear fuel upon contact with water is a central issue for the assessment of the safety of geological disposal concepts. Several studies have been conducted aiming at understanding matrix dissolution as well as the rapid/instant release of radionuclides separated from the UO 2 matrix; there are however questions remaining regarding how higher burn-up affects the amount of fission products segregated to the fuel-cladding gap and grain boundaries. In this study we have performed aerated fuel corrosion and leaching experiments using spent nuclear fuels with a range of burnups, from~52 to~70 MWd/KgU. The samples have been prepared in different manners in order to investigate the release of the segregated fission products from the gap and the grain boundaries. Thus, the fuel samples are prepared either as segments with the cladding attached, as fragments without cladding, or as milled powder. The fuel samples are leached in aqueous solution in aerated conditions and the radionuclide release is monitored during several contact periods for up to five years total leaching time. The results from the leaching of fuel segments reveal congruent release of e.g. europium and neodymium whereas the release of those elements was lower than the U-238 release for the fragment samples. Potential explanations for this are discussed in the paper. The results show that the release rates of elements segregated from the fuel matrix were in general found to be lower from segments samples as compared to fragment samples, which can be attributed to the closed fuel-cladding gap inhibiting the exposure of the gap inventory to water for the high burnup fuels used in this study. Leaching of fuel powder (78 MWd/kgU local BU) by simultaneous grinding and leaching showed a fractional release of Cs-137 and I-129 of 1.5% and 1.8% respectively.
Journal of Nuclear Materials, 2006
Pure UO 2 is often used as a model compound when studying reactions of importance in a future dee... more Pure UO 2 is often used as a model compound when studying reactions of importance in a future deep repository for spent nuclear fuel. The reactivity of pure UO 2 is not expected to be identical to the reactivity of the UO 2-matrix of spent nuclear fuel for several reasons. One reason is that the spent fuel, due to the content of radionuclides, is continuously being self-irradiated. The aim of this study is to investigate how irradiation of solid UO 2 surfaces affects their reactivity towards oxidants. The effect of irradiation (c or electrons) on the reaction between solid UO 2 and MnO À 4 in aqueous solutions containing carbonate has been studied. It was found that irradiation with high doses (>40 kGy) increased the reactivity of the UO 2 to about 1.3 times the reactivity of unirradiated UO 2 .
The general subject of this thesis is oxidative dissolution of UO 2. The dissolution of UO 2 is m... more The general subject of this thesis is oxidative dissolution of UO 2. The dissolution of UO 2 is mainly investigated because of the importance of the UO 2 matrix of spent nuclear fuel as a barrier against radionuclide release in a future deep repository. U(IV) is extremely insoluble under the reducing conditions prevalent in a deep repository, whereas U(VI) is more soluble. Hence, oxidation of the UO 2-matrix will affect its solubility and thereby its function as a barrier. In this thesis the relative efficiency of one-and two electron oxidants in dissolving UO 2 is studied. The oxidative dissolution yield of UO 2 was found to differ between one-and two-electron oxidants. At low oxidant concentrations the dissolution yields for one-electron oxidants are significantly lower than for two-electron oxidants. However, the dissolution yield for one-electron oxidants increases with increasing oxidant concentration, which could be rationalized by the increased probability for two consecutive one-electron oxidations at the same site and the increased possibility for disproportionation. This licentiate thesis is based on the following publications:
Journal of Nuclear Materials, 2019
The release of radionuclides from spent nuclear fuel upon contact with water is a central issue f... more The release of radionuclides from spent nuclear fuel upon contact with water is a central issue for the assessment of the safety of geological disposal concepts. Several studies have been conducted aiming at understanding matrix dissolution as well as the rapid/instant release of radionuclides separated from the UO 2 matrix; there are however questions remaining regarding how higher burn-up affects the amount of fission products segregated to the fuel-cladding gap and grain boundaries. In this study we have performed aerated fuel corrosion and leaching experiments using spent nuclear fuels with a range of burnups, from~52 to~70 MWd/KgU. The samples have been prepared in different manners in order to investigate the release of the segregated fission products from the gap and the grain boundaries. Thus, the fuel samples are prepared either as segments with the cladding attached, as fragments without cladding, or as milled powder. The fuel samples are leached in aqueous solution in aerated conditions and the radionuclide release is monitored during several contact periods for up to five years total leaching time. The results from the leaching of fuel segments reveal congruent release of e.g. europium and neodymium whereas the release of those elements was lower than the U-238 release for the fragment samples. Potential explanations for this are discussed in the paper. The results show that the release rates of elements segregated from the fuel matrix were in general found to be lower from segments samples as compared to fragment samples, which can be attributed to the closed fuel-cladding gap inhibiting the exposure of the gap inventory to water for the high burnup fuels used in this study. Leaching of fuel powder (78 MWd/kgU local BU) by simultaneous grinding and leaching showed a fractional release of Cs-137 and I-129 of 1.5% and 1.8% respectively.
Journal of Nuclear Materials, 2006
Pure UO 2 is often used as a model compound when studying reactions of importance in a future dee... more Pure UO 2 is often used as a model compound when studying reactions of importance in a future deep repository for spent nuclear fuel. The reactivity of pure UO 2 is not expected to be identical to the reactivity of the UO 2-matrix of spent nuclear fuel for several reasons. One reason is that the spent fuel, due to the content of radionuclides, is continuously being self-irradiated. The aim of this study is to investigate how irradiation of solid UO 2 surfaces affects their reactivity towards oxidants. The effect of irradiation (c or electrons) on the reaction between solid UO 2 and MnO À 4 in aqueous solutions containing carbonate has been studied. It was found that irradiation with high doses (>40 kGy) increased the reactivity of the UO 2 to about 1.3 times the reactivity of unirradiated UO 2 .