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.

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.

An engineering model of corrosion of zirconium-niobium alloys is described. It takes account of the alloying composition, the content of lithium and boron in the coolant, the heat flux on the surface of fuel elements and the intensity of the neutron irradiation. The parametric dependences used in the model are based on the results of tests performed in autoclaves and research reactors. The results of verification of the model on data from post-reactor studies of PWR and VVER fuel assemblies operating in nominal regimes are presented.The domestically produced zirconium-niobium alloys E-110 and E-635 are ordinarily used as structural materials in VVER fuel assemblies. The alloys M5 (Zr-1%Nb-Fe) and ZIRLO (Zr-1%Nb-Sn-Fe) are used in PWR fuel assemblies ( Table 1). The alloys E-110opt and E-635M based on zirconium sponge are under consideration as structural materials for Russian made fuel assemblies in the KVADRAT fuel assemblies.The main differences in water chemistry between PWR and VVER are the use of LiOH instead of KOH and dosing hydrogen instead ammonia in the coolant. Water chemistry based on LiOH can give rise to accelerated corrosion of zirconium alloys [2]. For this reason, research for additional validation of the corrosion behavior of the modified alloys E-110 and -635 is being conducted abroad for licensing fuel.Engineering models of the corrosion of zirconium alloys are used for validation of fuel serviceability [2-4]. These models are used to predict the thickness of the oxide layer and the hydrogen content in the cladding. In 2010-2012, an engineering model of the corrosion of E-110 and -635 alloys and their modifications was developed at the Bochvar All-Russia Research Institute for Inorganic Materials (VNIINM). The model is used in the computational fuel software making it possible to take account of the evolution of the height distribution of the temperature in the fuel-element cladding at the oxide-metal interface and the dependence of the corrosion rate on the heat flux density on the surface of fuel-element cladding. The present article describes the model and presents some results of verification.General Form of the Engineering Model. The engineering model was constructed using an approach similar to foreign models [2,3]. The stages of oxidation before and after the turning point are examined. The model is based on the Arrhenius relation for the corrosion rate in pure water as a function of temperature. Additional factors take account of irradiation, water chemistry and the concentration of the alloying additives. Each factor represents a parametric dependence,

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