M. Griffiths

A review of microstructure evolution in zirconium alloys during irradiation

Abstract The microstructure developed in Zr alloys during irradiation has proven to be increasingly complicated as specimens of varying composition and thermo-mechanical history have been irradiated to higher fluences. Cavities form in neutron and electron irradiated Zr at temperatures between 625 and 775 K. They are less common in the Zircaloys either because of vacancy trapping or because c- component dislocation loops compete for vacancies. There is a complex dependence on the purity of the material which is related to the presence of precipitates or interstitial solutes and on the presence of insoluble gases such as He. The latter is necessary to prevent the collapse of small vacancy clusters to form vacancy loops. Vacancy and interstitial dislocation loops having Burgers vectors of b = 1 3 〈11 2 0〉 ( 〈a〉 type) coexist during neutron and electron irradiation. They are generally aligned in rows or layers parallel with (0001). There is a marked decrease in the stability of vacancy 〈a〉 type loops at temperatures ⩾ 723K. Dislocation loops having Burgers vectors of b = 1 2 [0001] (vacancy type), b = 1 3 〈11 2 3〉 (interstitial type) and 1 6 〈20 2 3〉 (undetermined character) have been identified in Zr following electron irradiation at temperatures between 573–773 K. Vacancy loops of b = 1 6 〈20 2 3〉 have been observed in Zr and Zr alloys with a high interstitial solute content following neutron irradiation at temperatures between 560–773 K. Their formation and growth is determined by a combination of: (i) size effect interaction, dependent on their Burgers vector; (ii) stress, generated in collision cascades during neutron irradiation or in oxidised thin foils during electron irradiation; (iii) interstitial solute, which may lower the stacking-fault energy or segregate to the dislocations bounding the loops; and (iv) anisotropic interstitial diffusion (principally in the basal plane). The dislocation network in cold-worked materials is retained during neutron irradiation, at least for fluences up to 7 × 10 25 n m −2 at 573 K. Intermetallic precipitates of Zr with Fe, Cr and Ni undergo amorphous transformations (depending on the precipitate composition) during neutron irradiation at temperatures 560 K. The secondary precipitates are often distributed in rows or layers parallel with (0001). Radiation-enhanced precipitation from solution has been observed in ZrSn and ZrNb alloys at temperatures between about 573–873 K.

Phase instability, decomposition and redistribution of intermetallic precipitates in Zircaloy-2 and -4 during neutron irradiation

Intermetallic particles of Zr(Cr, Fe)2 and Zr2(Ni, Fe) in Zircaloy-2 and -4 undergo structural, chemical and morphological changes during neutron irradiation which are dependent on temperature, fluence and flux. At low temperatures (< 350 K), both particle types become amorphous. At intermediate temperatures (520–600 K), the Zr2(Ni, Fe) particles remain crystalline and the Zr(Cr, Fe)2 particles become amorphous. The susceptibility of the Zr(Cr, Fe)2 particles to the amorphous transformation increases with increasing flux and Cr content and is also dependent on the depletion of Fe and Cr below the stoichiometric limit. The volume of material transformed is proportional to the neutron fluence. At high temperatures (640–710 K) both the Zr2(Ni, Fe) and Zr(Cr, Fe)2 particles remain crystalline. For all temperatures there is radiation-induced dissolution of the intermetallic particles either as a result of sputtering or diffusion (possibly by an interstitial mechanism). At high temperatures (640–710 K) there is redistribution of the solute resulting in secondary precipitation in the matrix and at grain boundaries and a change in morphology of the intermetallic particles. Sn-rich precipitates are observed following irradiation to moderate fluences (~5 × 1025n m−2) at about 675 K and appear to be the result of radiation-enhanced diffusion.

The formation of c-component defects in zirconium alloys during neutron irradiation

Abstract There is a correlation between the existence of c- component defects and accelerated irradiation growth of annealed Zr and Zircaloy-2 and -4. Analysis shows that these defects are vacancy, basal plane dislocation loops having Burgers vectors of b = 1 6 〈2023〉. Their formation is related to the solute content of the Zr (either as an impurity or as an alloying element) and, in many cases, appears to be dependent on the dissolution of intermetallic particles during irradiation. This could account for why irradiation growth of annealed Zircaloys increases at high temperatures and why there is an incubation period for the onset of accelerated or “Breakaway” growth, i.e. because fluence and temperature are factors governing the extent of solute dissolution.

Evolution of microstructure in hcp metals during irradiation

The radiation damage microstructure in hexagonal-close-packed (hcp) metals is more complex than that of cubic materials, primarily because the crystal structure is different (hexagonal rather than cubic symmetry). For pure hcp metals the principal habit plane for dislocation loop nucleation is generally the most close-packed plane and this varies with c/a ratio; {1010} prism planes being the most close-packed in materials with c/a 1.732. Recent electron irradiation results indicate that, for hcp metals such as Zr and Mg, with a c/a < 1.732, there is a higher probability for c-component dislocation loops forming on basal planes (in addition to the prism plane loops) when the materials are impure. Both interstitial and vacancy loops are affected by impurities and, once nucleated, loop growth is often very rapid. One can conclude that the most important stage of c-component loop formation is nucleation, the probability of basal loop nucleation increasing with increasing impurity content.

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.

Grain boundary sinks in neutron-irradiated Zr and Zr-alloys

Abstract Samples of annealed sponge and crystal-bar Zr and Zircaloy-2 have been examined following irradiation in EBR-II at temperatures ~ 700 K. Loop analysis shows that there is selective denuding of interstitial loops near to some grain boundaries indicating that such boundaries are net sinks for interstitial point defects. Furthermore, in sponge Zr and Zircaloy-2, vacancy c- component loops are observed running into the grain boundaries showing that the grain boundaries are not preferred sinks for vacancies. Cavities are observed in all samples. In crystal-bar Zr and sponge Zr they are mostly observed adjacent to grain boundaries. They are also sometimes found within grains associated with precipitates. The cavities are more common in the crystal-bar Zr and this is probably because both the sponge Zr and Zircaloy-2 contain vacancy c- component loops which compete for vacancies (assuming that the cavities are vacancy sinks). Only some of the grain boundaries have cavities adjacent to them and this may be related to the orientation of the boundary.

HVEM study of the effects of alloying elements and impurities on radiation damage in Zr-alloys

Abstract The effect of alloying additions and impurities on radiation damage in Zr-alloys has been studied using a high voltage electron microscope (HVEM). The electron damage results have been compared with neutron damage results for the same materials. There is some qualtitative agreement between the two techniques in that the relative susceptibility to c -component loop formation is the same in both the neutron and electron case. Cavity formation, however, occurs more readily during electron irradiation. Because of the relationship between c -component loop formation and accelerated growth, these results show that electron irradiation using a HVEM can provide a preliminary indication of the susceptibility of Zr-alloys to accelerated growth prior to their use in nuclear reactors.

Microstructure Evolution in Zr Alloys during Irradiation: Dose, Dose Rate, and Impurity Dependence

The performance of zirconium alloys in BWR, PWR, and PHWR nuclear reactors is dependent on the microstructure. Accordingly, the characterization of the microstructure is an integral part of any study conducted to develop models for in-reactor performance. Although the as-fabricated microstructure (texture, grain size, dislocation density, and phase or precipitate distribution) determines the basic physical properties of a given component, there are changes that occur during irradiation that can have a significant effect on these properties. Microstructures that illustrate specific features of the radiation damage that forms in Zr alloys will be illustrated and discussed in terms of the dose, dose rate, and impurity factors that are applicable.

Mechanical Properties of Zr-2.5Nb Pressure Tubes Made from Electrolytic Powder

Pressure tubes in CANDU and RBMK reactors are fabricated from Zr-2.5Nb alloy. This paper describes the mechanical properties of tubes used in power reactors made by four routes using electrolytic powder as the base material. The microstructures developed by each route are distinct: CW: cold-worked material consists of flattened α-Zr grains surrounded by a skin of β-Zr phase; used in CANDU 6 reactors. CW-A: material that was cold-worked and annealed at 540°C contains elongated α-Zr grains mixed with equiaxed α-Zr grains and particles of β-Nb phase; used in all RBMK 1000 reactors. TMT-1: material quenched from the (α+β)-Zr phase into water follow by cold-working consists of α′-phase and between 10 and 20 % of untransformed α-phase; used in RBMK 1500, Ignalina 1. TMT-2: material quenched from the (α+β)-phase into argon-helium gas mixture followed by cold-working consists of Widmanstatten α-phase and untransformed α-phase. This material is used in RBMK 1500, Ignalina 2. The CW and TMT-2 tubes have a higher proportion of grains with basal plane normals in the transverse direction, FT of 0.52 to 0.57, than in the radial direction, FR of 0.38, while quenched and annealed materials (TMT-1 and CW-A) have similar values of FT and FR, about 0.38 in quenched materials and 0.41 in annealed materials. Transverse tensile strength, crack growth resistance, dJ/da, and axial crack velocity, VH, of delayed hydride cracking (DHC) were evaluated, using standard techniques, between 250 and 300°C on as-fabricated materials. In-reactor creep deformation was evaluated from measurements of tube diameter in RBMK 1000s, RBMK 1500s and two CANDU 6 power reactors. Strength and crack growth resistance were measured on TMT-1 and TMT-2 tubes removed from Ignalina NPP Units 1 and 2 after 12–17 years of in-reactor service. As-received cold-worked material had the highest strength; the annealed material had the lowest strength while the quenched materials had intermediate strength. Irradiation increased the strength by about 200 MPa in all four materials. Although DHC is sensitive to texture and the distribution of the β-Zr phase, the dominating factor controlling crack velocity appears to be material strength: with an increase of strength by a factor of about two, VH increased by a factor of 30. Since harmful trace elements were well controlled during manufacturing, other factors affecting crack growth resistance could be assessed. Again, strength appeared important; dJ/da declined approximately linearly with increase in strength induced by irradiation, decreasing from about 350 to 100 MPa as the strength increased from about 250 to 850 MPa. The exception was TMT-2 material where crack growth resistance was maintained after irradiation. TMT-2 material also had good diametral creep resistance in-reactor, attributed to both its texture and grain structure. The other three materials had similar creep resistance controlled mostly by their texture.

Damage Dependence of Irradiation Deformation of Zr-2.5Nb Pressure Tubes

The diametral expansion and elongation rates of Zr-2.5Nb pressure tubes in CANDU™ (CANada Deuterium Uranium) nuclear reactors are important properties that limit their useful life and the maximum power level for reactor operation. For a given set of operating conditions there is considerable variability in the deformation rates because of the variations in as-fabricated microstructure and chemistry from tube-to-tube — specifically grain size, crystallographic texture, and oxygen content. The as-fabricated microstructure also varies within a given tube, the largest variation occurring along the length, and this is a result of cooling of the tube during the extrusion process. During service in a nuclear reactor, the microstructure evolves further, and this additional change in microstructure is primarily dependent on the rate of radiation damage (determined by the fast neutron flux), the temperature, and the time. Both the fast neutron flux and temperature vary at all points within the pressure tube. For a given material microstructure, the deformation is a function of the operating conditions: coolant pressure (stress), temperature, and neutron flux. In principle, the deformation rate is a linear function of fast neutron flux, and this is mostly true for fast neutron fluxes of the order of 1017 n.m−2.s−1. Recent analyses of data from pressure tubes measured over long periods of operation in reactor have shown that the steady-state diametral creep rates are not linear with fast neutron flux for fluxes up to about 0.5 × 1017 n.m−2.s−1. A qualitative model has been developed to account for the observed behavior based on the modifying effects of neutron flux and temperature on the microstructure. The model describes the suppression of thermal creep and the transition from thermal to irradiation creep with increasing neutron flux.