The nuclear industry strives to reduce the fuel cycle cost, enhance flexibility and improve the reliability of operation. This can be done by both increasing the fuel weight and optimizing rod internal properties that affect operational margins. Further, there is focus on reducing the consequences of fuel failures. To meet these demands Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO 2 fuel containing additions of chromium and aluminium oxides. This paper presents results from the extensive investigation program which covered examinations of doped and reference standard pellets both in the manufactured and irradiated states.The additives facilitate pellet densification during sintering and enlarge the pellet grain size. The final manufactured doped pellets reach about 0.5% higher density within a shorter sintering time and a five fold larger grain size compared with standard UO 2 fuel pellets. The physical properties of the pellets, including heat capacity, thermal expansion coefficient, melting temperature, thermal diffusivity, have been investigated and differences between the doped and standard UO 2 pellets are small.The in-reactor performance of the ADOPT pellets has been investigated in pool-side and hotcell Post Irradiation Examinations (PIEs), as well as in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced fission gas release, improved PCI performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. Fuel segments have been exposed to ramp tests and enhanced power steady-state operation in the Studsvik R2 reactor after base-irradiation to above 30 MWd/kgU in a commercial BWR. ADOPT reveals up to 50% lower fission gas release than standard UO 2 pellets. The fuel degradation behaviour has been studied in two oxidizing tests, a thermal-microbalance test and an erosion test under irradiation. The tests show that ADOPT pellets have a reduced rate of fuel washout, as compared to standard UO 2 pellets.Fuel rods with ADOPT pellets have been irradiated in several light water reactors (LWRs) since 1999, including two full SVEA-96 Optima2 reloads in 2005.

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It is now well acknowledged that, after a prototypical loss of coolant accident (LOCA) transient, the resultant mechanical properties of fuel cladding tubes depend strongly on the oxygen content of the residual prior-β layer, as this phase is the only metallic part of the high-temperature oxidized cladding that may show some residual ductility. The aim of this study is to obtain relevant information on the evolution of the mechanical properties, on the one hand, of the prior-β structure as a function of the oxygen content, assuming that there is a critical oxygen content that leads to a ductile-to-brittle failure mode transition at low testing temperatures (20–135°C); and on the other hand, of the α(O) structure as a function of the oxygen content. Sheets of Zircaloy-4, 1 to 3 mm thick, and M5®M5® is a registered trademark of AREVA-NP. advanced alloys from AREVA NP have been studied. To obtain different oxygen contents, they were oxidized at high temperature and then annealed under vacuum in order to reduce the oxide layer. Systematic post-treatment measurements of the oxygen concentration and of its homogeneity within the sheet thickness were performed. The different prior-β and α(O) structures thus obtained have homogeneous oxygen content between ∼0.14 wt. % and 0.9 wt. % and ∼2 wt. % and 7 wt. %, respectively. Such oxygen concentration ranges cover the solubility values that are expected in the β phase and in the α(O) phase at high temperatures typical of LOCA transients. Detailed microstructure investigations were subsequently performed on the prior-β structures since it is considered to be the most important layer when regarding the post-quench mechanical behavior of the material. Continuous cooling temperature (CCT) phase diagrams as a function of the oxygen content were established to correctly interpret the results. Electron backscattered diffraction (EBSD) analysis has then allowed the crystallographic orientations and the morphology of prior-β phase sub-grains to be determined. For each considered prior-β grain, it was possible to interpret the data by taking into account the “Bürgers” crystallographic relationship between the parent β phase and the resultant α phase. Complementary electron probe microanalysis (EPMA) was also used. These last experiments have shown a spatial fluctuation of the oxygen content within the microstructure that depends both on the nominal oxygen content and on the cooling rate. Nanohardness measurements were also performed and correlated with this oxygen spatial partition. These measurements proved to be useful for the understanding of the tensile macroscopic mechanical behavior. Finally, on the one hand, tensile tests were performed on prior-β phase at testing temperatures ranging from −100°C up to 260°C. The ductile-to-brittle temperature transition and the mechanical constitutive laws as a function of the oxygen content were then described. These tests show the existence of a ductile-to-brittle failure mode transition at 20°C for a critical oxygen concentration of ∼0.5 wt. %. A detailed fractographic analysis was performed to assess the failure mechanism. On the other hand, four-point bending tests were conducted on α(O) phase at 25°C and 135°C in order to obtain behavior laws. Preliminary finite element calculations were performed to simulate ring compression tests carried out on multi-layered high-temperature oxidized cladding tubes.