On account of the search for the optimal composition and structure-phase state of Zr alloys much attention is paid to upgrade the E110 (Zr-1 %Nb) and E635 (Zr-1 %Nb-0.35 %Fe-1.2 %Sn) alloys that have proved well in terms of irradiation-induced creep and growth, high strength characteristics, and corrosion. The difference between the alloy properties is determined by their states related to their compositions. The structure-phase state of the Zr-Nb and Zr-Nb-Fe-Sn systems has been studied after heat treatment in the α-- and α + β- regions and its influence on the irradiation-induced growth (IIG) during BOR-60 irradiation at T =315–350 °C was investigated. A substantial difference has been shown in the deformation effected by IIG of those alloys, it is less for Zr-Nb-Fe-Sn alloys in dissimilar structure-phase states. The incubation period of the accelerated growth stage is determined by the α-matrix composition, the phase state and the initial dislocation structure. Neutron irradiation leads to a redistribution of alloying elements between the matrix and the precipitates, and to changes in the α-solid solution composition. These changes affect accumulation and mobility of irradiation defects, anisotropy and formation of vacancy c-component dislocation loops. The appearance of c-loops usually correlates with an axial direction acceleration of the IIG of tubes conforming to their texture. The basic regularities of the phase transformation have been established: a) β-Nb precipitates in Zr-Nb alloys are altered in composition to reduce the Nb content from 85–90 % to ∼ 50 %, fine precipitates likely enriched in Nb are formed; b) β-Zr precipitates are subject to irradiation-stimulated decomposition; c) Laves phase precipitates change composition (the content of Fe decreases) and crystal structure, HCP to BCC (β-Nb); d) (Zr,Nb)2Fe precipitates having the FCC lattice retain their composition and crystal structure; e) no amorphization of any secondary phase precipitates is observable under the given conditions of irradiation (T = 315–350 °C). Based on the dpa, the results were compared pertaining to Zr-alloy IIG deformation vs. fluence in various reactors at different energies of fast neutrons. The presented graphs enable comparison between the results of numerous experiments and enable predictions of Zr-material behavior in long-term operation and at high burn-up in commercial reactors.

Studied were evolution of dislocation structure, phase, and element composition of binary alloys Zr-1Nb and Zr-2.5Nb and multicomponent alloys Zr-1Nb-1.2Sn-0.4Fe and Zr-1.2Sn-0.4Fe under neutron irradiation. The investigations were carried out using cladding and pressure tubes before and after irradiation to a fluence of ∼1026 n/m2 (E ≥ 0.1 MeV) in experimental and commercial reactors at 300 to 350°C using TEM, EDX, and XRD. In most cases, irradiation-induced defects are in the form of dislocation loops with Burgers vector 1/3 ⟨1120⟩. The density of dislocations with a ⟨c⟩ component is less than 2 × 1014 m-2. A higher fluence or the presence of strain results in the ordering of the dislocation structure of ⟨c⟩ component and ⟨a⟩-type dislocation loops. Before irradiation, the multicomponent alloys contain fine precipitates of Zr-Nb-Fe composition, and the matrix is depleted in Fe. Under irradiation, recrystallization proceeds intensively (as distinct from Zr-Nb alloys), changes take place in size, distribution, and composition of precipitates (with a relative decrease of Fe content compared to Nb), and the Fecontent of α-Zr matrix is increased. None of the materials studied showed any significant evidence of secondary phase particle amorphization. The density of dislocations with ⟨a⟩ and ⟨c⟩ components and irradiation-induced defects, their mean size, the extent of ordering, and the planes of their occurrence were determined. A comparison was made between irradiation-induced evolutions of microstructures of the different alloys.

The studies of the dislocation structure, phase, and microchemical compositions of alloy Zr-1Nb-1.2Sn-0.35Fe (E635) and its modifications containing Fe from 0.15 to 0.65% were carried out before and after research reactor irradiation at ∼350°C to maximal fluence of ∼1027 m-2 (E > 0.1 MeV) and at ∼60°C. The size and concentration of the <a>-type loops depend on the alloy composition and fluence and saturate even at low doses (<1 dpa). The evolution of the <c>-component dislocation structure in recrystallized alloys of E365 type is determined by the chemical and phase compositions of alloys specifically, by the Fe/Nb ratio and the threshold dose, and is consistent with the irradiation growth strain acceleration. In E635 alloy containing 0.15%Fe the accelerated growth is observed after the dose of 15 dpa and is attended with the evolution of the <c> dislocation structure which is similar to Zr-1Nb (E110) alloy behavior. The irradiation induced growth of E635 type alloy containing 0.65% Fe is similar to that of E635 having the normal composition; no <c> dislocations are observed up to the dose of 20 dpa. E635 alloy contains precipitates Zr(Nb1-xFex)2 (HCP) as the basic excess phase and individual (Zr,Nb)2Fe (FCC) precipitates; in 0.15%Fe alloy aside from Zr(Nb,Fe)2 also β-Nb (BCC) particles precipitate, while 0.65%Fe alloy contains Zr(Nb,Fe)2 and (Zr,Nb)2Fe particles. Irradiation at 330 –350 °C does not effect an amorphization of β-Nb or Zr(Nb,Fe)2 precipitates; however, at higher fluences the β-Nb phase becomes depleted in Nb and Zr(Nb,Fe)2 in Fe. Irradiation at 60°C leads to the amorphization of Zr(Nb,Fe)2 in E635. The analysis revealed that the key factors promoting a delay in the accelerated irradiation growth in Zr-Nb-Fe-Sn alloys are the composition of (Nb,Fe,Sn) solid solution and the Fe/Nb ratio in alloys.

In the search for more optimal core materials for a water cooled reactor at extended burnup, much attention is paid to alloys of the Zr-Nb and Zr-Nb-Fe-Sn systems. E110 and E635 alloys are two such. In the current VVER fuel cycle, the E110 alloy is used as fuel cladding and in SG components. The E635 alloy is under development as a fuel cladding and for fuel assembly structural elements for water cooled reactors of the VVER and RBMK types. E110, while having a unique corrosion resistance in pressurized water reactors, is subject to noticeable disadvantages in terms of corrosion resistance under conditions of boiling and higher coolant oxygen contents as well as in deformation stability under stresses and irradiation. Currently, the E635 alloy has passed the most important steps of qualification and is being introduced into cores as a material for guide thimbles, central tubes, and stiff frame angles in VVER-1000 FAA and FA-2. Properties of alloys are governed by their compositions and microstructure and even small changes in composition (Nb, Fe, Sn) and processing (heating in the α or the α+β regions) lead to substantial changes in properties as a result of changes in second phase precipitates and matrix composition. ATEM was used to study structure—phase states of a series of alloys Zr-(0.6–1.2) Nb-(0–0.6) Fe-(0–1.5) Sn (% weight), to determine the microstructural characteristics of recrystallized cladding tubes and the temperature stability regions of β-Nb, β-Zr, Zr(Nb,Fe)2, and (Zr,Nb)2Fe second phase precipitates. An increase in the relative content of iron R=Fe/(Fe+Nb) results in a larger volume fraction of (Zr,Nb)2 Fe precipitates. β-Nb and Zr(Nb,Fe)2 particles are completely dissolved at ⩽750°C, the (Zr,Nb)2Fe phase at ⩽800°C. Autoclave corrosion tests revealed that the corrosion resistance of the materials depends on alloy composition. The content of tin lowered down to 0.8 % reduces weight gains in water, water containing Li, and particularly in steam. The content of Nb reduced to 0.6 % results in lower weight gains in water and steam and higher weight gains in Li containing water. The optimal content of iron in Zr-Nb-Fe-Sn alloys for corrosion resistance depends on the R ratio and makes up 0.2–0.4 %. Tests of samples produced from tubes of the above alloys and irradiated in BOR-60 at 315–345°C show that alloying Zr-Nb alloys with iron and tin improves their resistance to irradiation growth and creep. Sn and a higher Fe content in solid solution effected by transfer of Fe from the Laves phase precipitates to the matrix under irradiation strengthens the alloys. The influence of irradiation on phase compositions was established using irradiated samples (gas filled and unstressed) of cladding tubes: β-Nb (85–90 % Nb) precipitates become depleted in niobium (or enriched in zirconium) to 50–60 % Nb and finely dispersed irradiation induced second particles (IIPs) enriched in niobium are formed. The Laves phase becomes depleted in iron and alters its crystal structure from hcp to bcc of the β-Nb type. The fcc (Zr,Nb)2Fe precipitates retain on the whole their composition and structure, but the peripheries of particles reveal structural features, possibly related to niobium redistribution. No amorphization of any of the precipitates was identified. Alloy composition and applied stress under irradiation influence density and distribution of dislocation loops and IIP precipitates. Proceeding from results of out-of-pile and from post-irradiation examinations of the structure and properties of E110 and E635 type cladding tubes, compositions of alloys having improved corrosion and irradiation resistances are proposed. E110 type (Zr-1Nb-0.1Fe-0.1O) alloy features enhanced strength characteristics as a result of iron transfer from Laves phase precipitates to the matrix under irradiation, lower irradiation induced growth strain, and irradiation-thermal creep. An E635 type alloy (tin and niobium content lowered down to <0.8 %) has a higher corrosion resistance and comparable creep and growth resistance as compared to the standard E635 alloy.

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