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Small angle neutron scattering (SANS) results are presented for Linde 80 welds irradiated, as part of the B&W Owners Group Integrated Surveillance Program, at low fluxes (< 1015 n/m2-s) to fluences from 0.29 to 3.5x1023 n/m2 (E > 1 MeV) at irradiation temperatures from 276 to 292°C. The welds all contain about 0.6 Ni (all composition units are in wt.%), 0.009 to 0.18 P and 0.05 to 0.28 Cu. In the welds with significant amounts of copper (> 0.2 Cu) the measured defect scattering cross sections were consistent with either: a) copper rich precipitates (CRPs) alloyed with manganese and nickel; or b) dominant CRP scattering, plus a weak contribution from so-called matrix defect features. Similar weak scattering was observed in a low copper (0.06 Cu) weld. The identity of matrix defect features cannot be determined from the SANS data alone, but the scattering is consistent with the presence of subnanometer vacancy cluster-solute complexes. The general character of the CRPs, and the trends in their number density, volume fraction and average radius as a function of fluence and irradiation temperature, are very similar to those observed in a wide range of pressure vessel-type steels irradiated in test reactors at intermediate to high flux. The SANS data in the surveillance welds is also in unity with: a) thermodynamic-kinetic radiation enhanced diffusion models of CRP evolution; b) mechanical property changes, including predictions of the correlations of the surveillance data base; and c) an atomic scale, atom probe field ion microscopy study into the nanostructure-chemistry of a CRP.

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Thermomechanical processing of Zircaloy-4 cladding plays a major role in determining its deformation behavior. Crystallographic texture and the related anisotropy in mechanical properties of Zircaloy-4 have been shown to be affected by different processing paths. In this program, the deformation behavior of four Zircaloy-4 cladding types was evaluated in laboratory and in-reactor studies under typical pressurized-water reactor (PWR) conditions. In particular, the creep behavior, stress-free growth, and mechanical property changes of these materials were examined. The irradiation program consisted of four specimen cluster assemblies, each containing 16 rods, irradiated in guide tube locations of fuel assemblies in the Oconee-2 reactor. Twelve of the 16 rods in each cluster were prepressurized so that nominal inreactor compressive hoop stresses of 69.0, 86.2, and 103.4 MPa (10, 12.5, and 15 ksi) were maintained. The remaining four rods were open to the coolant and served as stress-free irradiation growth specimens. One specimen cluster was examined after each of four reactor cycles. Local fast neutron fluences ranged up to 1.3 × 1022 n/cm2 (E > 0.1 MeV), while irradiation temperatures ranged from 569 to 591 K (565 to 605°F). Post-irradiation tension and burst tests were conducted at 616 K (650°F) to determine changes in the mechanical properties as a function of fluence. The ex-reactor thermal creep behavior was examined at 672 K (750°F) in a series of multiaxial creep tests where the hoop/axial stress ratio was held constant at several different values. These data were used to construct a partial creep locus at a constant value of the dissipative work function. Results indicate that the in-reactor axial creep and stress-free growth are highly dependent on the cladding fabrication schedules. Most of the axial strain exhibited by the cold-worked, stress-relieved cladding types S1, V1, and V2, consists of irradiation growth, while in the recrystallized type S2, axial creep predominates. In-reactor diametral creep strain exhibited an initial transient regime at fast neutron fluences below 3 to 5 × 1021n/cm2 (E > 0.1 MeV) with a creep rate decreasing to a steady state at higher fluences. In contrast to axial creep behavior, the cladding with the highest level of retained cold work (S1) exhibited the highest diametral creepdown while recrystallized type S2 showed the least creepdown of the four cladding types. The hoop/axial strain rate ratio, εθ/εz, remained essentially constant over the fluence and stress ranges investigated, but its magnitude was significantly different between the S1 and S2 cladding types. A comparison between the in-reactor deformation and ex-reactor thermal creep behavior shows that the relative ranking of the materials in creep resistance is consistent in both environments. An analysis of the anisotropy of in-reactor deformation and a comparison with the thermal creep anisotropy are also provided. Post-irradiation axial tension and burst tests at 616 K (650°F) indicated that the most cold-worked material, S1, maintained the highest strength. In axial tension, it also retained the highest uniform and total elongations. In the burst tests, the more recrystallized cladding types. S2 and V2, exhibited the highest total circumferential elongations. Irradiation-induced changes in mechanical properties were significantly greater in the recrystallized cladding material.

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