Effect of Cr on the creep properties of zirconium alloys

To investigate the effect of the Cr element in zirconium-based alloys on the creep properties, Zr–1.2Nb–0.1Cr, Zr–1.2Nb–0.5Cr, Zr–1.2Nb–0.3Sn–0.1Cr, and Zr–1.2Nb–0.2Sn–0.3Cr alloys were manufactured and creep tested under a constant stress of 120 MPa at 380 °C for 250 days. As the amount of Cr as well as Sn increased in the studied alloys, the creep strain rates decreased. The strengthening effect of Cr is considered to be efficient when the zirconium alloy contains Nb as an alloying element. The relative contribution of Cr against Sn contents on creep resistance was also observed to be comparable.     

Measurements of hydrogen permeation and absorption in zirconium oxide scales

The hydrogen absorption and the permeation behavior through the oxide layer formed on modified Zircaloy-4 (Zry-4) alloys were investigated. The modified Zry-4 was prepared by altering the chemical composition of standard Zry-4. The tin content of Zry-4 (1.5 wt%) was reduced to 0.5 wt%, and alloying elements Si, O and Nb were added from 0.01 to 0.2 wt%. The oxide layers were grown in a static autoclave at 360 °C under 18.3 MPa for 150 days. Fick’s law was used to calculate the diffusivity of hydrogen after the steady state of the permeation flux was reached. The diffusivity of hydrogen in the 0.5Sn–0.1Nb–0.1Fe–0.2Cr–0.2O–Zr specimen was lower than that in the 1.5Sn–0.2Fe–0.1Cr–0.1O–0.01Si–Zr and Zry-4 specimens. As the area fraction of precipitates increased, the hydrogen diffusivity increased whereas an inverse relationship between the diffusivity and the amount of the tetragonal phase was observed. In addition to the oxide structural study, the effects of the microstructure of the zirconium alloys such as precipitates and grain boundaries on the hydrogen absorption were studied.     

Effects of molybdenum on microstructural evolution and mechanical properties in Zr–Nb alloys as nuclear fuel cladding materials

The Zr–Nb alloys were modified by doping of Mo as a minor alloying element to seek for the nuclear fuel cladding materials with better characteristics. The effects of Mo on microstructural evolution and mechanical properties in Zr–Nb alloys were systematically investigated and elucidated. Results showed that the martensitic microstructure, a mixture of lath martensites and lens martensites with internal twins, was observed in the alloys quenched from β-phase. Width of the lath martensite reduced with the increasing Mo concentration, and the volume fraction of lens martensite increased with increase in the Mo concentration. After final annealing, a new kind of precipitate, namely β-(Nb, Mo, Zr), was identified in the Mo-containing alloys. It was also found that Mo reduced the growth of the precipitates but increased their number density. Furthermore, Mo addition retarded the recrystallization process strongly and reduced the grain size significantly. In terms of the mechanical properties, Mo addition enhanced the yield strength and the ultimate tensile strength at room temperature, however decreased the ductility. The grain size strengthening was presumed as the greatest contributor in this system.     

High temperature oxidation of some zirconium alloys in flowing oxygen

The oxidation behavior of three zirconium alloys, Zr-2.2wt%Hf, Zr-2.5wt%Nb and Zr-3wt%Nb-1wt%Sn, has been studied in flowing oxygen in the temperature range 873–1173 K to 120 ks (2000 min). Zr-2.5Nb and Zr-3Nb-1Sn showed a transition to rapid linear kinetics after initial parabolic oxidation at all temperatures. Zr-2.2Hf, on the other hand, showed this transition at temperatures in the range 973–1173 K; no transition was observed at 873 K within the oxidation times reported. Zr-2.2Hf showed the smallest weight gains, followed by Zr-2.5Nb and Zr-3Nb-1Sn. Increased oxidation rates and shorter time-to-rate transition of Zr-2.5Nb and Zr-3Nb-1Sn as compared with Zr-2.2Hf are attributed to the presence of the alloying elements Nb, Sn and Hf. Based on the Nomura-Akutsu model, Hf should delay the rate transition, while Nb and Sn lead to shorter transition times. The scale on Zr-2.2Hf was identified as monoclinic zirconia, while the tetragonal phase, 6ZrO₂ · Nb₂O₅, was contained in the monoclinic zirconia scales on both other alloys.     

Characteristics of the thermal desorption of helium from the model alloys Kh16N15 and Kh13

Thermal desorption spectrometry is used to study the characteristic behavior of helium in the new model bcc and fcc alloys Fe-13Cr and Fe-16Cr-15Ni, which are the basis of structural reactor steels, with the substitution elements Mo, W, Nb, Ta, V, and Ti and carbon. It is found that the gas-release spectra of bcc alloys are more complicated. The activation energy of gas release is calculated. It is shown that the alloying elements do not have a unique effect on the mechanisms of helium release from bcc and fcc alloys. The data obtained are discussed from the standpoint of the formation and thermal stability of different helium—vacancy complexes and the effect of doping by a change of the diffusion mechanisms on the growth and migration of bubbles. __________ Translated from Atomnaya énergiya, Vol. 104, No. 1, pp. 13–17, January, 2008.     

Effect of alloying elements (Cu, Fe, and Nb) on the creep properties of Zr alloys

The thermal creep behaviors of Zr-based alloys containing Cu, Fe and Nb were investigated under constant load stress at temperatures of 280 and 330 °C, and a stress range of 100-140 MPa. To evaluate an alloying effect on a creep, Zr-based alloys were selected as the binary and ternary systems of Zr-0.3Cu, Zr-0.3Fe, Zr-0.5Nb-0.3Cu and Zr-0.5Nb-0.3Fe. The final annealing of these alloys was performed at 510 °C for 8 h to obtain a recrystallization structure for all the tested alloys. A microstructure characterization test was carried out for the samples before and after the creep test by using TEM, and the results were used to understand the creep mechanism. Creep tests were performed for up to 70 h, which showed a steady-state secondary creep rate in all the alloys. The value of the stress exponent was about 5.5 in all the alloys. The dislocation density was increased by increasing the applied stress, regardless of the alloy system. From the results of this study, it was revealed that the Nb as an alloying element showed the strongest effect on the creep resistance among the added alloying elements, and Fe was more effective than Cu from the viewpoint of creep resistance.     

Oxidation behaviour of zirconium alloys and their precipitates – A mechanistic study

The precipitate oxidation behaviour of binary zirconium alloys containing 1 wt.% Fe, Ni, Cr or 0.6 wt.% Nb was characterised in TEM on FIB prepared transverse sections of the oxide and reported in previous studies [1], [2]. In the present study the following alloys: Zr1%Cu, Zr0.5%Cu0.5%Mo and pure Zr are analysed to add to the available information. In all cases, the observed precipitate oxidation behaviour in the oxide close to the metal-oxide interface could be described either with delayed oxidation with respect to the matrix or simultaneous oxidation as the surrounding zirconium matrix. Attempt was made to explain these observations, with different parameters such as precipitate size and structure, composition and thermodynamic properties. It was concluded that the thermodynamics with the new approach presented could explain most precisely their behaviour, considering the precipitate stoichiometry and the free energy of oxidation of the constituting elements.The surface topography of the oxidised materials, as well as the microstructure of the oxide presenting microcracks have been examined. A systematic presence of microcracks above the precipitates exhibiting delayed oxidation has been found; the height of these crack calculated using the Pilling–Bedworth ratios of different phases present, can explain their origin. The protrusions at the surface in the case of materials containing large precipitates can be unambiguously correlated to the presence of these latter, and the height can be correlated to the Pilling–Bedworth ratios of the phases present as well as the diffusion of the alloying elements to the surface and their subsequent oxidation. This latter behaviour was much more considerable in the case of Fe and Cu with Fe showing systematically diffusion to the outer surface.     

Effect of impurity and alloying elements on Zr grain boundary strength from first-principles computations

Effects of twenty impurity and alloy elements on the strength of a Zr(0 0 0 1)/Zr(0 0 0 1) ∑7 twist grain boundary were studied using a first-principles density functional approach. A ranking in the order of most weakening to most strengthening was: Cs, I, He, Te, Sb, Li, O, Sn, Cd, H, Si, C, N, B, U, Ni, Hf, Nb, Cr, and Fe. Segregation energies for these elements to the grain boundary and the Zr(0 0 0 1) surface were also calculated. Calculations showed that the weakening grain boundary elements He, I, and Cs have a strong driving force for segregation to the grain boundary from bulk Zr. Zircaloy cladding failures (pellet-clad interactions) in commercial fuel systems and separate effects test results provide context for these computational results.     

In situ oxidation of zirconium binary alloys by environmental SEM and analysis by AFM, FIB, and TEM

Binary Zr-alloys containing 1%Fe and 1% Ni (large precipitates) and 1% Cr and 0.6% Nb (small precipitates), as well as a pure Zr sample were exposed in situ at 130 Pa water vapour pressure at 415 °C in an environmental SEM. The surface topography and composition of each sample was characterised before in situ experiments, during and after oxidation. After oxidation the surface was characterised by SEM and EDS, AFM and TEM combined with EDS. Focused ion beam was used to prepare cross sections of the metal-oxide interface and for the preparation of TEM thin foils.The oxidation behaviour of precipitates for these alloying elements can be characterised into two large families, those which show a rapid oxidation and those which induce a delayed oxidation in comparison with the Zr-matrix.At 415 °C after 1 h of oxidation for Zr1%Fe and Zr1%Ni, the formation of protrusions could be detected at the surface, being related to underlying SPP in the oxide. On Zr1%Cr and Zr0.6%Nb unoxidised SPPs were observed in the oxide, close to the metal-oxide interface. These SPPs were, however, oxidised close to the outer surface of the oxide. The surface roughness was increased for all materials after in situ oxidation, however, only for Zr1%Fe and Zr1%Ni protrusions appeared on the surface during oxidation. It was subsequently demonstrated that these latter correspond to the position of SPPs. For Zr1%Fe the surface roughness increased more than in the other materials and on these protrusions small iron oxide crystals have been observed at the surface. These observations confirm that Fe has a different behaviour compared to the other SPP forming elements, and it diffuses out to the free surface of the material.These alloying elements being the constituents of the commercial alloys (Fe and Cr for Zircaloy-4; Fe, Cr and Ni for Zircaloy-2 and Nb for all Nb-containing alloys), this study allows to separate their individual influence and can allow a subsequent comparison to the behaviour of those more complex alloys.     

Radiation-induced segregation in titanium alloys

Radiation-induced segregation (RIS) of V, Mo, Nb, Ta, Zr, and Sn in binary titanium alloys was investigated to test the solute size effect correlation in hcp alloys. Undersize Mo segregates weakly toward the sinks. Nb and Ta, which are slightly oversize in Ti, undergo little or no RIS. Oversize Zr solute in Ti segregates away from the sinks, whereas undersize Ti solute in Zr is enriched at sinks. All of these results are in accord with the solute size effect correlation. Surprisingly, Sn, which is significantly oversize in Ti, appears to segregate very little. The postirradiation annealing of Ti-3V and Ti-8Al-1V-1Mo confirmed that segregation of undersize V toward sinks is radiation-induced. Measurements of temperature and dose dependence in binary and complex alloys showed that the degree of V segregation has a maximum at ~ 600 °C and obeys parabolic growth kinetics in its early stages but probably saturates at a rather low dose ( ~ 0.8 dpa).     

Effect of alloying elements on proclivity of zirconium alloys for nodular corrosion

An approach to assessing the effect of alloying elements on the proclivity of zirconium alloys for nodular corrosion is described. The approach is based on an analysis of the stability of the corrosion front. The results of a simplified analysis for iron and nickel additives to Zircaloy-2 and -4 are in agreement with the experimental data. The approach described can also be used for zirconium-niobium alloys, but this requires taking account of the mutual effect of the atoms of niobium and other alloying substances. For detailed quantitative assessments of the effect of alloying elements on the proclivity of zirconium alloys for nodular corrosion, the stability analysis must be expanded and including the use of self-consistent numerical models which take account of radiation effects and oxygen transfer through the oxide film. Translated from Atomnaya énergiya, Vol. 106, No. 2, pp. 94–99, February, 2009.     

The Effect of Oxide Microstructure on Kinetic Transition in Out-of-pile Steam Corrosion Test for Zircaloy-2 and Nb-added Zircaloy-2

In order to study the mechanism of kinetic transition of corrosion rate for zirconium alloys, oxide films formed on Zircaloy-2 (Zry-2) and Nb-added Zircaloy-2 (0.5Nb/Zry-2) in steam at 673 K and 10.3 MPa were examined with TEM and SIMS.

Kinetic transition occurred at almost the same oxide thicknesses for both Zry-2 and 0.5Nb/Zry-2, but the corrosion rate after the transitions were quite different for the two alloys. Zircaloy-2 showed cyclical oxidation, while the weight gain of 0.5Nb/Zry-2 increased linearly.

The morphology and crystal structure were similar for the oxides of the two alloys and both the oxide films still mainly consisted of columnar grains even after the transition. Interface layers which mainly consisted of a-Zr crystallites were observed for both alloys and the oxygen content in the interface layers increased after the transition.

The solute concentrations of Fe, Cr and Ni became higher, accompanying the increase of oxygen concentrations at columnar grain boundaries in the oxide films after the transition for 0.5Nb/Zry-2. It was thought that the properties of grain boundaries of the 0.5Nb/Zry-2 oxide films changed after the transition, and the increase in oxygen diffusivity at grain boundaries caused the linear increase in weight gain.     

Performance of FCCI barrier foils for U-Zr-X metallic fuel

Diffusion couple tests of U-Zr or U-Zr-Ce alloys vs. ferritic martensitic steels such as HT9 or T91 were carried out in order to evaluate the performance of the diffusion barrier candidates. Elemental metal foils of Zr, Nb, Ti, Mo, Ta, V and Cr were very effective in inhibiting interdiffusion between these fuels and steels. Eutectic melting between the fuels and steels was not observed in any of the diffusion couples using these diffusion barrier foils at annealing temperatures up to 800 °C. Among the metallic foils evaluated in this study, V and Cr exhibited the most promising performances as a diffusion barrier material for eliminating the fuel cladding chemical interaction problem. However, Zr, Nb and Ti showed an active interaction with the fuel mainly due to the large U solubility.     

Effect of V and Nb on precipitation behavior and mechanical properties of high Cr steel

Employment of high Cr steel as a main structural material is considered as a way to achieve economical competitiveness of fast breeder reactors. V and Nb are believed to improve the high temperature strength of high Cr steels by precipitating as carbides and/or nitrides, namely fine MX particles. However, the long term efficiency and stability of such precipitation strengthening mechanisms provided by the fine MX particles have not been clarified yet. A series of trial products controlling V and Nb contents is produced and mechanical tests are conducted to investigate the effect of these elements on the mechanical properties and the long term stability of the MX strengthening mechanism. Before and after a long term aging process, metallurgical examinations and quantitative analyses are conducted to investigate the effect of these elements on microstructure evolutions. Based on these results, the long term efficiency and stability of the strengthening mechanisms provided by the fine MX particles are discussed. Higher strength and lower ductility are obtained with the increases of V and Nb contents, although the influence of Nb content tends to be saturated at about 0.01 mass%. MX does not grow and any new precipitates cannot be observed after aging at 873 K for 6000 h. Therefore, it is expected that MX is stable after aging at 823 K for approximately 167,000 h based on Larson-Miller parameter.     

Irradiation-Induced Dissolution of Precipitates in Zircaloy-2

Calibration curves of extremely low concentrations of the alloying elements Sn, Fe, Cr and Ni in Zircaloy were obtained, using standard samples, by energy dispersive X-ray spectroscopy to measure concentration distributions of alloying elements dissolved in the Zircaloy matrix. Their detectable limits were 0.21 at% for Sn, 0.06 at% for Fe. 0.04 at% for Cr and 0.03 at% for Ni. Then concentration distributions of alloying elements in unirradiated and neutron irradiated Zircaloy-2 were measured using these calibration curves. It was confirmed that neutron irradiation increased the dissolved concentrations of Fe. Cr and Ni. Furthermore, Cr diffused slower than Fe and Ni. It was suggested that the rate limiting process of irradiation-induced dissolution from Fe, Cr-type precipitates into the matrix was the diffusion of alloying atoms in the precipitates and that the dissolution process proceeded due to displacement of alloying atoms from the precipitates into the matrix and diffusion in the matrix.     

Creep behavior of niobium-modified zirconium alloys

Zirconium (Zr) alloys remain as the main cladding materials in most water reactors. Historically, a series of Zircaloys were developed, and two versions, Zircaloy-2 and -4, are still employed in many reactors. The recent trend is to use the Nb-modified zirconium alloys as the Nb addition improves cladding performance in various ways, most significant being superior long term corrosion resistance. Hence, new alloys with Nb additions have recently been developed, such as Zirlo^™2 and M5^™3. Although it is known that creep properties improve, there have been very few data available to precisely evaluate the creep characteristics of new commercial alloys. However, the creep behavior of many Nb-modified zirconium alloys has been studied in several occasions. In this study, we have collected the creep data of these Nb-modified alloys from the open literature as well as our own study over a wide range of stresses and temperatures. The data have been compared with those of conventional Zr and Zircaloys to determine the exact role Nb plays. It has been argued that Nb-modified zirconium alloys would behave as Class-A alloys (stress exponent of 3) with the Nb atoms forming solute atmospheres around dislocations and thus, impeding dislocation glide under suitable conditions. On the other hand, zirconium and Zircaloys behave as Class-M alloys with a stress exponent of ?4, attesting to the dislocation climb-controlled deformation mode.     

Post-irradiation ductility loss of FeNibase precipitation hardenable alloys

Several γ′- and γ′/γ″-strengthened Fe-Ni-base superalloys have shown near-zero ductility after neutron irradiation to fluences of $\text{～ 4 × 10}{}^{22}\text{n/cm}{}^2$, $\text{E > 0.1 MeV}$, at 500 to 650°C. The ductility loss is most pronounced in solution-treated or in solution-treated and aged specimens tested at 110°C above the irradiation test temperature. Failed specimens exhibit brittle intergranular fractures. Microstructural examination of the embrittled specimens showed that continuous or semi-continuous coatings of γ′ formed at grain boundaries during the irradiation. In some cases, the grain boundaries were also decorated by small bubbles, thought to be transmutation-induced helium or contained trace elements such as sulphur and phosphorus. All of these grain boundary alterations are attributed to radiation-induced solute segregation. Microanalyses of the γ′ coatings indicate that Ni, Al, Si, Ti and Nb had segregated to grain boundary sinks in irradiated FeNiCr based alloys. Nonequilibrium segregation of helium and trace impurities is also considered likely. The role of radiation-induced segregation in the embrittlement phenomenon is consistent with the observation that introducing a high density of dislocation sinks by cold-working reduces γ′ formation at the grain boundaries and reduces the ductility loss. The embrittlement is attributed to concurrent strengthening within grains by irradiation-induced γ′ precipitation and brittle cleavage failure of grain boundary precipitates.     

Formation of Gas Pores in Nickel Alloys and Structural Steel under Irradiation by Helium Ions

Transmission electron microscopy is used to study the development of helium porosity in binary alloys of nickel with elements possessing a different dimensional atomic mismatch with nickel – from negative (beryllium and silicon) to positive (molybdenum, tungsten, aluminum, titanium, tantalum, tin, and zirconium), in structural steels ChS-68, ÉP-150, and the nickel alloy KhNM. The gas pores were produced by irradiation with 40 keV He⁺ up to fluence 5·10²⁰ m^–2 at 650 and 20°C followed by annealing at 650°C for 1 h. It is shown that under high-temperature annealing beryllium and silicon, relative to nickel, give rise to the formation of larger bubbles, while elements with a larger positive size mismatch with nickel atoms substantially decrease the size and increase the density of the bubbles. On the whole, as atomic radius and the concentration of the alloying element increases in alloys, the gas swelling of the irradiated layer decreases. Under post-irradiation annealing, bubbles with the largest diameter and the lowest density develop in nickel. Any alloying used decreases the size and increases the density of bubbles. The data obtained are discussed from the standpoint of the formation of various vacancy complexes of helium and their thermal stability.     

Effect of phosphorus on the mechanical behavior of a Zr-Nb alloy

The flow stress of Zr-1.0Nb increased with the addition of 20 ppm phosphorous and the activation volume decreased with the addition of 300 ppm phosphorous at room temperature. The rate-controlling mechanism of the deformation of Zr-Nb-P is thought to be the dislocation-solute interaction in which the segregation of alloying elements affects the activation length. Less effective strengthening of phosphorous in Zr-Nb compared to sulphur was explained by a smaller electrostatic interaction between phosphorous atoms and dislocations.     

Transmission electron microscopy examination of oxide layers formed on Zr alloys

A transmission electron microscopy investigation was performed on oxides formed on three zirconium alloys (Zircaloy-4, ZIRLO and Zr-2.5Nb) in pure water and lithiated water environments. This research is part of a systematic study of oxide microstructures using various techniques to explain differences in corrosion rates of different zirconium alloys. In this work, cross-sectional transmission electron microscopy was used to determine the morphology of the oxide layers (grain size and shape, oxide phases, texture, cracks, and incorporation of precipitates). These characteristics were found to vary with the alloy chemistry, the corrosion environment, and the distance from the oxide/metal interface. These are discussed and used in conjunction with observations from other techniques to derive a mechanism of oxide growth in zirconium alloys.