Zircaloy-4 and Incoloy 800H/HT Alloys for the Current and Future Nuclear Fuel Claddings

Abstract

Fuel cladding is one of the most critical components of nuclear reactors; so it is important to improve our understanding of various properties and behaviors of the cladding under different conditions approximating the nuclear reactor environment. Moreover, the efficiency of energy production, in addition to safety concerns, has resulted in progressive improvement of nuclear reactors design from Generation I to Generation IV. To complement this progressive trend, materials used for fuel cladding need to be improved or new materials should be developed. In this thesis, I address problems in the improvement of present fuel cladding and also investigate fuel cladding materials to be used in future Generation IV nuclear reactors. In the case of current Zircaloy-4 fuel claddings, a detailed evaluation of the surface roughness effects on their performance and properties of Zircaloy-4 fuel claddings was studied. A smoother surface on Zircaloy-4 cladding tubes is demanded by the customers; however no systematic study is available addressing the effect of surface roughness on the claddings’ performance. Thus the effects of surface roughness on texture, oxidation, hydriding behaviors and mechanical properties of Zircaloy-4 cladding tubes were investigated using various methods. It was found that surface roughness has some effects on the oxidation of Zircaloy-4. Increasing the surface roughness would increase the weight gain, however, this effect was more pronounced at the initial oxidation stages. Synchrotron techniques were used to characterize the electronic structure of zirconium alloys in their oxidized and hydrided states. With this approach, complex interactions between hydrogen and oxygen in the zirconium matrix could be investigated, which could not be resolved using conventional methods. As a candidate for future fuel cladding material, Incoloy 800H/HT, which is expected to be considered in super-critical water-cooled Gen IV reactors, was studied in order to optimize microstructure, texture and grain boundary characteristics. A specific Thermo-Mechanical Processing (TMP) was employed to manipulate the texture, microstructure and grain boundary character distribution. The deformation and annealing textures of thermo-mechanically processed samples were investigated by means of X-ray diffraction and orientation imaging microscopy. It was found that different rolling paths lead to different textures. The origin of different textures in differently (unidirectional and cross) rolled Incoloy 800H/HT at high deformation strains were investigated. In addition, the recrystallization kinetic of differently rolled samples was studied. It was found that the oriented nucleation plays an important role in determining the recrystallization texture. Unidirectional rolled samples exhibited a faster recrystallization kinetic compared with cross rolled ones, due to the presence of γ-fibre. The effect of the aforementioned microstructural parameters (grain size, texture and GBCD) on the oxidation resistance of Incoloy 800H/HT in super-critical water was investigated. It was found that the oxidation resistance of Incoloy 800H/HT can be improved by TMP. The optimum TMP process for enhancing the oxidation resistance was proposed. Microstructural parameters that can improve the oxidation resistance of Incoloy 800H/HT were identified. These findings will contribute to the effective selection of fuel cladding material for application in Gen IV SCW reactors.