Ronald Bert Adamson

Characterization of neutron irradiation damage in zirconium alloys — an international “round-robin” experiment

The results are given of an international “round-robin” experiment to study the nature of the damage structure in neutron irradiated zirconium and zircaloy-2 using transmission electron microscopy. The damage structure consists entirely of 13α dislocation loops and no evidence has been found for c-component loops. Both vacancy and interstitial loops were found in specimens irradiated at 400 °C, with an excess of vacancy loops. Quantitative measurements of loop size distributions and loop concentrations are reported. All specimens exhibited “corduroy” contrast to varying degrees. The importance of choice of imaging conditions to minimize the contrast from thin foil artefacts such as oxide films and surface hydrides is stressed. The significance of the results is briefly discussed with reference to current theories of irradiation growth.

Precipitates in zircaloy: Identification and the effects of irradiation and thermal treatment

The effects of thermal treatment and neutron irradiation on precipitates in Zircaloy have been investigated by analytical electron microscopy. Intermetallics Zr(Fe, Cr)2 and Zr2(Fe, Ni) are the major precipitates in α-recrystallized Zircaloy-2 and Zr(Fe, Cr)2 is the major precipitate in Zircaloy-4. Zr-silicide and Zr-Cu-sulfide are minor precipitates and amorphous inclusions, respectively. In Zircaloy heat treated in the α + β- or β-phase fields, fine platelet Zr4(Fe, Cr) precipitates decorate the boundaries of lamellae formed on quenching. Neutron irradiation results in a change from a crystalline to an amorphous structure in the Zr(Fe, Cr)2 precipitates. The amorphous transformation starts at the periphery of the precipitate and the thickness of the amorphous ring increases as the fluence increases. The process leading to amorphization is discussed based on the effects of irradiation-induced point defects. In α + β- or β-heat-treated Zircalioy, the fine Zr4(Fe, Cr) precipitates are completely dissolved into the matrix upon irradiation.

Neutron damage in zirconium alloys irradiated at 644 to 710 k

The microstructure of annealed crystal-bar zirconium, sponge zirconium and Zircaloy-2 and -4 have been analysed following neutron irradiation in EBR II over the temperature range of 644–710 K for neutron fluences up to 6– x 1025 n m−2 (E>1 MeV). There is a correlation between measured high irradiation growth strains and the existence of vacancy c-component dislocation loops. The concentration of these faulted 16〈2023〉 dislocation loops is highest in alloyed or impure Zr. There is dissolution of Fe, Cr and Ni from intermetallic particles during irradiation. The amount of solute dissolution and secondary precipitation is dependent on the irradiation temperature and fluence and is most widespread for the Zircaloy irradiated to high fluences at high temperatures. Sn-rich precipitates are also observed in the Zircaloys and are the result of radiation-enhanced diffusion.

Effects of microchemistry and precipitate size on nodular corrosion resistance of Zircaloy-2

Abstract Atom probe studies have been performed on various specimens of Zircaloy-2 in order to measure the concentrations of the alloying elements Cr, Fe, and Ni in the α-Zr matrix. The concentrations were compared with the densities of nodules and corrosion weight gains obtained in two-step (683 K, 793 K) steam tests. The data partially confirm the earlier hypothesis that corrosion is lowered by higher concentrations of solutes in the matrix. However, TEM measurements clearly show that small precipitate size is also an important factor in minimizing corrosion. Optimum corrosion resistance is achieved when the solute contents of the α-Zr matrix are high and the precipitates are small.

Implications of Zircaloy creep and growth to light water reactor performance

Deformation of zirconium alloy components in nuclear reactors has been a concern since the decision of Admiral Rickover to use them in the US Navy submarine reactors. With the exception of the first few light water reactors (LWRs) most of the core structural materials have been fabricated from either Zircaloy-2 or Zircaloy-4. Performance of these alloys has been extremely good, even though the effects of irradiation on deformation magnitudes and mechanisms were not fully appreciated until extensive service and in-reactor tests were accomplished. Since the reactor components are designed to operate at stress levels well below yield for normal conditions, the only significant deformation is time dependent. Although creep was anticipated, the enhancement by neutron irradiation and the stress-free, nearly constant-volume shape change known as irradiation growth were not known prior to materials testing in reactors under controlled conditions. Both of these phenomena have significant impact on performance and must be accounted for properly in design. Although irradiation creep and growth have resulted in only one significant performance problem (creep collapse of fuel cladding, which has been eliminated), deformation magnitudes and, particularly, differentials in strain magnitudes, are a continuing source of interest. Factors that affect dimensional stability due to both creep and growth include temperature, fluence, residual stress, texture, and microstructure. The first two are reactor variables and the others are related to component fabrication history. This paper includes a review of the applications of Zircaloy creep and growth to LWR fuel designs, a review of the impact of in-reactor creep and growth on fuel rod and fuel assembly performance, and comments on potential improvements. Since the reactor design, fuel design and the core environment in BWRs and PWRs are quite different, appropriate separation of the application of effects are made; of course, the basic phenomena are the same in both systems.

Precipitate behavior in zirconium-based alloys in BWRs

Abstract The metallic elements Cr, Fe, Ni, Nb, and Mo have low solubilities in zirconium, so that zirconium-based alloys made with these elements contain numerous small precipitates. Some of these undergo substantial changes in composition, crystalline structure, and size during in-reactor exposure. Data on these phenomenon have been obtained using the scanning transmission electron microscope (STEM) to study precipitates before and after reactor exposure. Results on the following alloys are reported here: Zircaloy-2 (Zr-1.5Sn-0.15Fe-0.10Cr-0.05Ni), Zry-0.2Nb (Zircaloy-2 with 0.2Nb), NoCr (Zr-1.5Sn-0.17Fe-0.17Ni), NSF-2 (Zr-1Nb-1Sn-0.2Fe), and XLL (Zr-1.5Sn-0.3Nb-0.3Mo). The dissolution of Fe from various precipitates is apparent from direct compositional measurements. The dissolution of Cr can be inferred in Zry-0.2Nb from Cr:Nb gradients in amorphous Cr-Fe-Zr-Nb precipitates and from the formation of intergranular, Cr,Fe-rich precipitates during postirradiation annealing. Two types of phase change have been seen. First, crystalline phases can become amorphous, as with hexagonal Laves Zr(Fe,Cr)2 in Zircaloy-2 and cubic Fe(Zr,Nb)2 in NSF-2. In the second type, long-range order is eliminated, as with Laves (Mo,Nb)2Zr precipitates in XLL which become BCC. The precipitate size distributions can change in reactor. The greatest changes were measured in small, Cr,Fe-rich precipitates.

High-Temperature Irradiation Growth in Zircaloy

Irradiation growth behavior of Zircaloy-2 and -4 was studied on specimens irradiated in an experimental breeder reactor. Measurements on Zircaloy-2 slab materials provided evidence that growth was a constant volume process up to 680 K. The growth strains were shown to be determined by the crystallographic texture. The growth strains for annealed and cold-worked Zircaloy were strongly dependent on irradiation temperature and varied linearly with fluence. The data suggest a transition from saturating steady-state growth at lower temperatures to increasing and eventually high steady-state rates at higher temperatures in the temperature range used (644-723K). 29 refs.

The influence of localized stresses and strains on the iodine scc behavior of Zircaloy-2 tubing

Abstract The segmented expanding mandrel test is modified to provide additional information on the localized stresses and strains during ramp-hold tests. A chamfered specimen is designed with four flats on the outer diameter for the purpose of creating sectors of variable wall thickness which establishes a more nearly plane strain state in the thinnest-gauge sections. Ramp-hold tests conducted in 40 MPa of iodine at 325°C show that at a measured diametral strain of 0.8%, chamfered specimens fail while regular tubing specimens require diametral strains of 2% or greater for failure. Calculations made with a two-dimensional, finite elements code show that at a measured diametral strain of 0.8%, the local strains in the clad ID over the segment interfaces are nearly identical for the regular and chamfered samples while the local stress in the chamfered sample is 15% greater. Results suggest that the smaller diametral failure strain of the chamfered specimen at low measured diametral strains is mainly due to its near plane strain condition and the higher hoop stresses resulting therefrom.