Recrystallisation behaviour of a fully austenitic Nb-stabilised stainless steel

We have performed an in-depth characterisation of the microstructure evolution of 20Cr-25Ni Nb-stabilised austenitic stainless steel during 1 h isochronal annealing up to 1100°C using scanning electron microscopy. This steel grade is used as cladding material in Advanced Gas-cooled fission reactors, due to its resistance to thermal creep and oxidation. The initial deformed microstructure undergoes recrystallisation via a strain-induced boundary migration mechanism, attaining a fully recrystallised microstructure at 850°C. A number of twins are observed in the vicinity of deformation bands prior to the start of recrystallisation. New Nb(C, N) particles form gradually in the microstructure, and the particle dispersion presents a maximum volume fraction of 2.7% at 930°C. At higher temperatures, the smaller particles become unstable and gradually dissolve in the matrix. Consequently, the Zener pinning pressure exerted on the grain boundaries is progressively released, triggering the growth of the austenite grains up to an average size of ∼47 μm at 1100°C. The observed temperature window for recrystallisation and grain growth can be predicted by a unified model based primarily on the migration of high- and low-angle grain boundaries. Lay Description: Austenitic stainless steel containing high percentage of chromium and nickel is currently used as fuel cladding material in the British Advanced Gas-cooled Reactors (AGR). This material has been chosen because of its high resistance to thermal creep and corrosion, both enhanced by the presence of a fine dispersion of carbonitrides precipitated during the cladding thermomechanical processing. During the time spent in the reactor core, few fuel cladding elements can become susceptible to local chromium depletion at grain boundaries, which is ascribed to the time evolution of the microstructural damage caused by the neutron bombardment in the reactor core. This depletion might increase the susceptibility of this steel to intergranular corrosion attacks during medium-to-long term storage of spent fuel elements in water ponds. The severity of the local chromium depletion depends not only on the irradiation conditions, but also on the grain boundary geometry. We have investigated the recovery, recrystallisation and grain growth of AGR stainless steel during 1 h annealing at selected temperatures relevant for the thermomechanical processing of the steel claddings, focusing on the formation and evolution of grain boundaries and second phases. These two features play a key role in the progression of the neutron damage and the subsequent development of local chromium depletion during reactor service operations. A deep understanding of the mechanisms and conditions behind their formation during the thermomechanical processing of the cladding material and their interaction with each other constitutes the foundation to evaluate, and potentially mitigate, the effect of irradiation on the cladding material.</p

The Effect of Iron on Dislocation Evolution in Model and Commercial Zirconium Alloys

Although the evolution of irradiation-induced dislocation loops has been well correlated with irradiation-induced growth phenomena, the effect of alloying elements on this evolution remains elusive, especially at low fluences. To develop a more mechanistic understanding of the role iron has on loop formation, we used state-of-the-art techniques to study a proton-irradiated Zr-0.1Fe alloy and proton- and neutron-irradiated Zircaloy-2. The two alloys were irradiated with 2-MeV protons up to 7 dpa at 350\ub0C and Zircaloy-2 up to 14.7 7 1025n • m-2, approximately 24 dpa, in a boiling water reactor at approximately 300\ub0C. Baseline transmission electron microscopy showed that the Zr3Fe secondary-phase particles in the binary system were larger and fewer in number than the Zr (Fe, Cr)2and Zr2(Fe, Ni) particles in Zircaloy-2. An analysis of the irradiated binary alloy revealed only limited dissolution of Ze3Fe, suggesting little dispersion of iron into the matrix, while at the same time a higher 〈a〉-loop density was observed compared with Zircaloy-2 at equivalent proton dose levels. We also found that the redistribution of iron during irradiation led to the formation of iron nanoclusters. A delay in the onset of 〈c〉-loop nucleation in proton-irradiated Zircaloy-2 compared with the binary alloy was observed. The effect of iron redistributed from secondary-phase particles because of dissolution on the density and morphology of 〈a〉 and 〈c〉 loops is described. The implication this may have on irradiation-induced growth of zirconium fuel cladding is also discussed

Isothermal annealing behaviour of nuclear grade 20Cr-25Ni austenitic stainless steel

We have performed an in-depth characterisation of the microstructure evolution of 20Cr-25Ni Nb-stabilised austenitic stainless steel during isothermal annealing at 930 °C using scanning and transmission electron microscopy. This steel grade is used as cladding material in advanced gas-cooled fission reactors, due to its resistance to thermal creep and water corrosion. The initial deformed microstructure undergoes recrystallisation via a strain-induced boundary migration mechanism, attaining a fully recrystallised microstructure after 120 s of annealing. The transition from low-to-high grain boundaries has already occurred after 15 s, together with an increase in the cube grain orientation at the expense of the S texture component. After 120 s, the grain boundary migration induces the formation of new fine Nb(C,N) particles, whereas the pre-existing particles become enriched in Ni and Si. The resulting particle population limits the grain growth in the austenitic matrix, based on the Zener pinning model, resulting in relatively small recrystallised austenite grains and a high density of high-angle and special coincidence-lattice-site grain boundaries, together with a large number of particle/matrix interfaces.</p

Local chemical instabilities in 20Cr-25Ni Nb-stabilised austenitic stainless steel induced by proton irradiation

We have assessed the local solute redistribution at defect sinks in 20Cr–25Ni Nb-stabilised austenitic stainless steel after proton irradiation at three temperatures, i.e. 420, 460 and 500 °C, up to a maximum damage level of 0.8 dpa. This material is currently being used as cladding in Advanced Gas-cooled Reactors (AGR), and potential local Cr depletions would compromise its resistance to intergranular corrosion attack during wet storage of spent fuel elements. Irradiation induces the depletion of Cr, Fe and, to a lesser extent, Mn from grain boundaries, whereas Ni and Si become enriched at those locations. The elemental profiles are symmetric and primarily W-shaped at 420 °C, whereas at higher temperatures asymmetric and double-peaked profiles are also detected, most likely as a result of grain boundary migration. High-angle grain boundaries with a misorientation angle ≥40° become mobile at 460 °C and especially at 500 °C, and also experience a relatively large solute redistribution, with local Cr contents in a significant number of boundaries falling below 12 wt% and profile widths ≥100 nm. However, coincidence site lattice boundaries (CSL) Σ3 boundaries prove to be resistant to Cr depletion and to boundary mobility. Local elemental patterns at radiation-induced dislocations seem to mimic those at grain boundaries, but do not trigger the formation of Ni 3 Si precipitates. Additionally, Ni and Si form a shell-like structure around the pre-existing Nb(C,N) precipitates, potentially leading to the transition into G-phase at higher damage levels.</p