Irradiation of Zircaloy affects its microstructure and macroscopical properties, for example, influencing its irradiation growth. To gain more insight into these phenomena, experimental fuel rods and growth specimens with various fabrication parameters were irradiated in a pressurized water reactor (PWR) to high fluences. Some of the growth specimens were exposed to a fast neutron fluence of up to 2.3 × 1022 cm-2 (⩾0.82 MeV) over a period of 10 years. Following exposure, the irradiation-induced alterations of the microstructure and the intermetallic precipitates were studied by optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM). At a temperature of 300°C during irradiation to fluences up to 7 × 1021 cm-2, growth increases with increasing yield strength. Recrystallized material, which has a low yield strength, exhibits an increased growth rate at very high fluences (⩾1 × 1022 cm-2). Postirradiation annealing studies indicate that the early irradiation growth of the recrystallized material can be recovered, whereas the later accelerated growth does not seem to be recoverable. At temperatures in the range of 330 to 350°C, the growth depends on the grain size, especially below 2 μm, and on the carbon-content. The effect of yield strength on growth was less at 330 to 350°C than at 300°C, probably because of an irradiation-induced recovery. Moreover, TEM showed that an irradiation-induced formation of dislocations with a c-component occurs at neutron fluences ⩾9 × 1021 cm-2, whereas dislocation loops or fine precipitates or both form at lower fluences. The intermetallic precipitates observed in the microstructure of the unirradiated, initial material exhibit two types of intermetallics. One type contains (Fe + Cr + Zr) and the other type contains only (Fe + Zr). The effect of irradiation on those intermetallics depends on temperature. At ⩽300°C the (Fe + Zr)-type intermetallic precipitate dissolves at neutron fluences above 5 × 1021 cm-2. The (Fe + Cr + Zr)-type precipitates become more and more amorphous and release iron to the matrix resulting in a decreasing Fe/Cr ratio. The diameter and the number of the precipitates decrease with increasing neutron fluences at this temperature. Only a few small precipitates can still be observed after a neutron fluence of 1.5 × 1022 cm-2. At temperatures above 340°C the size of intermetallics increases because of irradiation enhanced ripening.
The deformability of austenitic stainless steels and nickel-base alloys was studied in an experimental program combining the influence of irradiation in a light water reactor (LWR) core environment and high stresses and strains. Tubular specimens filled with ceramic mandrels were inserted into fuel elements of both a BWR and a PWR where they were exposed to the neutron flux and coolant water of the core. The ceramic mandrels swelled under irradiation and applied high stresses to the cladding tube specimens; hence, in case of sensitive material, this led to intergranular stress corrosion cracking. The strain was varied by choosing different ceramic materials and by the axial position of the specimens in the reactor core. During the refueling shutdown of the reactor, the specimens were examined by integrity testing and diameter measurements. The experimental setup, irradiation conditions, examination methods, and deduced stresses will be described and the status of the experiments outlined. Many materials failed by brittle cracking under the conditions applied. In the first program phase only a low phosphorus and silicon stainless steel and an Inconel alloy 718 with a special heat treatment were found to be resistant. In the second phase, when other material charges were used to verify the first results, the good performance of alloy 718 was confirmed. High purity austenitic stainless steels, however, failed during Phase 2 at the same low strain level as commercial purity material. In the case of alloy X-750, it was found that the material surface condition had a significant influence on the resistance to stress corrosion cracking.
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