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- Hansson, Niklas, 1992, et al.
(author)
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Geometrical aspects of alpha dose rates from UO2 based fuels
- 2022
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In: Radiation Physics and Chemistry. - : Elsevier BV. - 0969-806X .- 1879-0895. ; 199, s. 110336-
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Journal article (peer-reviewed)abstract
- Models for calculating dose rates of spherical particles as well as in fuel cracks are important for radionuclide source term estimations. Dose rates from UO2 based fuels were calculated for planar, spherical, and crack geometries. The escape probability for a-particles in spherical UO2 particles was derived as a continuous equation. The dose rate increased with increasing spherical radius due to the decreasing relative volume of the surrounding water layer. The model produced escape probabilities that were closely predicted by the theoretical derivation. It was shown that the dose rate in water filled fuel cracks with width smaller than the range of an alpha-particle led to dissolution rates that were lower per unit surface area.
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| 2. |
- Hansson, Niklas, 1992
(author)
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Oxidative dissolution of UO2 by α-radiolysis
- 2022
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Doctoral thesis (other academic/artistic)abstract
- To prevent the spread of radiotoxic nuclides in the environment, spent nuclear fuel generated by decades of nuclear power operation must be safely stored for at least 100 000 years. The KBS-3 method is a highly developed deep geological repository concept and is the first final repository design for high-level nuclear waste to be constructed. It contains a number of engineered barriers designed to prevent groundwater from coming into contact with the spent nuclear fuel. However, the consequences of groundwater coming into contact with the fuel must be considered when assessing the safety of this repository concept. After ~1000 years, the initially dominant γ-emitting elements have largely decayed, and the α-emitters dominate the radiation field. At the fuel-water interface, the fuel’s strong α-radiation field causes extensive radiolysis, creating locally oxidizing conditions. The oxidants formed can cause oxidation of the UO2 matrix from the U(IV) state to the U(VI) state, significantly increasing its solubility in the process. The water intrusion also leads to anoxic corrosion of the iron inserts, forming large amounts of H2 in the process. This process has been shown to protect nuclear fuel against oxidative dissolution. The oxidative dissolution of UO2-based materials has been experimentally studied and modelled in this work. Oxidation and dissolution of UO2 pellets were studied under an external irradiation source, in both Ar and H2 atmospheres. In the Ar atmosphere, the oxidation of UO2 was shown to take place through the incorporation of a significant U(V) oxidation state fraction. In the H2 atmosphere, the surface was protected during exposure to the external irradiation source against both surface oxidation and dissolution. Very low dissolution yields were found in the study of SIMFUEL, with H2 catalytically activated on the pellet surface, efficiently causing catalytic decomposition of H2O2 without leading to oxidative dissolution of the UO2 matrix. Highly Pu-doped MOX pellets showed a strong oxidative dissolution in the Ar atmosphere. This was somewhat mitigated in the D2 atmosphere. The modelled data were shown to accurately replicate the experimental results. Dissolved U(VI) was shown to be strongly reductively precipitated on corroding iron foils under anoxic conditions. This decreased the initially dissolved concentrations by three orders of magnitude over relatively short periods. This work furthers the understanding of oxidative dissolution of UO2-based materials under α-radiation fields and the effect of reducing agents present in the canister.
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