Effect of fission products on air-oxidation of LWR spent fuel

A detailed study of UO2 to U3O8 oxidation phases and the associated rate-limiting steps

“…Typically, a-U 3 O 7 is considered as the normal result of the oxidation at temperatures less than 200°C [3], whereas b-U 3 O 7 appears at temperatures higher than 200°C [3,7]. It is noted that some studies report intermediate tetragonal phases with c/a ratios ranging between those of a-U 3 O 7 and b-U 3 O 7 [6,9]. Finally, the oxidation of powders proceeds only from b-U 3 O 7 to U 3 O 8 , and not from a-U 3 O 7 [7].…”

“…(1), McEachern and Taylor used the term U 4 O 9 /U 3 O 7 to refer to the product of the first stage of air oxidation of UO 2 [3]. Several intermediate oxidation products have been reported, namely UO 2+x [4,5], U 4 O 9 [5], c-U 4 O 9 [6], a-U 3 O 7 [5,7], b-U 3 O 7 [5,7] and c-U 2 O 5 [8], whose crystallographic structures are not well established. These crystalline structures are derivatives of UO 2 .…”

Computational study of the energetics and defect clustering tendencies for Y- and La-doped UO2

“…trivalent) fission product dopant-oxygen vacancy clusters to the decreased rate of oxidative dissolution with increasing burnup in spent fuel [14]. The defect clusters are believed to reduce the concentration of mobile oxygen vacancies which facilitate oxidation and eventual dissolution of the fuel, in a manner similar to that by which rare earth doping impedes air oxidation of the cubic fluorite structure to orthorhombic U 3 O 8 [15][16][17][18]. Oxidation and dissolution of nuclear fuel in aqueous environments is of particular concern at the back end of the fuel cycle, i.e.…”

Oxidation behaviour of uranium in the internally gelated urania–ceria solid solutions – XRD and XPS studies

“…The effect of dopants on the rate of oxidation of UO 2 has been studied [15,16] to gain insight into the different reactivities of UO 2 and spent fuel. Thomas et al [17] found that UO 2 doped with 4-8% gadolinia oxidized at around 450°C to a cubic phase resembling U 4 O 9 and at around 575°C to U 3 O 8 . A similar behaviour was observed during the oxidation of light water reactor spent fuel.…”