Hydride Precipitation Research Articles

On the Precipitation of Hydrides in Metals under Stress - Finite Deformation Approach

An analytic relation for the terminal solid solubility of hydrogen in a stressed metal is derived, based on finite deformation theory. Phase transformation is assumed to be a reversible process, which occurs under local chemical equi- librium among hydride, metal and hydrogen in solid solution. Hydrogen terminal solid solubility depends on stress due to the interaction of the applied stress field with the field of the expanding hydride as well as due to the reduction of hydro- gen chemical potential in solid solution, caused by hydrostatic stress. The present analysis is complementary to hydride- induced embrittlement studies in metals, under conditions, which require the use of finite deformations.

Ta/Zr接合界面における水素拡散および水素化物析出の計算機シミュレーション

The concentration of hydrogen and precipitation of zirconium hydrides in Ta/Zr explosive bonded joint were analysed by computer simulation. Numerical model of hydride precipitation under hydrogen diffusion was simplified by the alternate model coupled the macroscopic hydrogen diffusion with the microscopic hydride precipitation. Effects of the initial hydrogen content in Ta, working degree of Zr and post-bond heat treatment on the hydrogen diffusion and hydride precipitation were investigated. Hydrogen was rapidly diffused from Ta substrate into Zr after explosive bonding and temporarily concentrated at Ta/Zr bond interface. Zirconium hydrides were precipitated and grew at Ta/Zr bond interface, and the precipitation zone of hydrides was enlarged with the lapse of time. The precipitation of zirconium hydrides was promoted when the initial hydrogen content in Ta and working degree of Zr were increased. The concentration of hydrogen and precipitation of hydrides at the bond interface were reduced and diminished by post-bond heat treatment at 373K. It was deduced that hydrogen embrittlement in Ta/Zr explosive bonded joint was caused by the precipitation of zirconium hydrides and concentration of hydrogen at Ta/Zr bond interface during the diffusion of hydrogen containing in Ta substrate.

Ta/Zr接合界面における水素脆化割れの特徴とZr母材の水素脆化機構

In order to clarify the behaviour and mechanism of the hydrogen embrittlement in SUS304ULC/Ta/Zr explosive bonded joint, the hydrogen embrittlement cracking at Ta/Zr bond interface was characterised. Cracks occurred in the Zr substrate along the wavy interface of the hydrogen-charged Ta/Zr joint. The cracking susceptibility increased drastically when the potential of specimen during hydrogen-charging was reduced below the redox potential of hydrogen. γ-ZrH and δ-ZrH were precipitated in the hydrogen-charged Zr and the precipitated γ-ZrH possessed a (0002)α-Zr//(11-1)γ-ZrH, [21-1-0]α-Zr//[110]γ-ZrH crystallographic relationship. An in-situ observation of the hydrogen embrittlement cracking with SEM and TEM revealed that cracks were initiated in zirconium hydrides and propagated preferentially along zirconium hydrides. These results suggested that the hydrogen embrittlement mechanism of the Zr base metal was caused by the precipitation of zirconium hydrides and the brittle fracture of them.

水中研磨過程でのTa/Zr界面の水素脆化割れ現象と水素脆化機構

Hydrogen embrittlement cracking behaviours of SUS304/Ta/Zr explosive bonded joint during underwater polishing were investigated. Hydrogen embrittlement cracks occurred in the Zr substrate adjacent to the Ta/Zr bond interface during underwater polishing. The open circuit potential of Zr during underwater polishing was drastically reduced immediately after mechanical polishing (within a fraction of a second). The hydrogen yields of Zr-Ta alloys and cold-worked Zr during underwater polishing were estimated from the corrosion current determined by the Tafel extrapolation method. The hydrogen yield increased with a decrease in the Ta content of Zr-Ta alloy, and with an increase in the degree of working (rolling reduction) of Zr. It was deduced that the mechanical grinding in water removing the passive oxide film on the Zr substrate led to the hydrogen absorption into the Zr substrate and the precipitation of zirconium hydrides. Accordingly, hydrogen embrittlement cracks occurred in the deformation layer of Zr around the Ta/Zr bond interface due to the tensile residual stress in the explosive bonded joint.

Ta/Zr接合界面における水素脆化に及ぼす諸因子の影響

In order to clarify the hydrogen embrittlement mechanism at Ta/Zr explosive bonded interface, the effects of the Ta content and microstructure on changes in hardness and toughness during hydrogen charging were investigated by using Zr-Ta binary alloys. Zirconium hydrides were precipitated after hydrogen charging in Zr-Ta alloys with the specific crystallographic relationship to α-Zr substrate, and there was no significant difference in the precipitation behaviour of zirconium hydrides with varying the Ta content and microstructure (α, α' and ωphase structures). The impact absorbed energies of Zr-Ta alloys were drastically reduced with an increase in hydrogen charging time, while their hardness slightly increased with hydrogen charging. A pure Zr was highly sensitive to hydrogen embrittlement, whereas the toughness value after hydrogen-charging was lowest when alloys involved ω phase structure, and it decreased with an increase in the Ta content of the alloy. Furthermore, the sensitivity of hydrogen embrittlement of the Zr substrate was heightened with an increase in the degree of cold working (rolling reduction). The hydrogen embrittlement crack occurring in Zr adjacent to the Ta/Zr bond interface would be attributed to the mechanism that cracks were preferentially initiated and propagated in the precipitated zirconium hydrides which were precipitated on the deformation texture of α-Zr (0001) formed during explosive bonding.

Effect of Nb on hydride embrittlement of Zr– xNb alloys

The effects of Nb on hydride Zr alloys were investigated. Various Zr alloys with different Nb content up to 2.0% were prepared in a sheet shape and charged with hydrogen up to 850 ppm. It was found that the fraction of recrystallized grains was reduced with increasing Nb content during the heat treatment. Intergrain and intragrain hydrides were tangled in unalloyed Zr, which has fully recrystallized grains. On the other hand, Nb containing Zr alloys that have partially cold-worked grains had only intergrain hydrides precipitated along the rolling direction. In tensile tests, elongation was decreased significantly with increasing hydrogen content for unalloyed Zr. However, the reduction of elongation was less significant for Nb-containing alloys. The softening of the Zr matrix was caused by hydrogen, which was arrested in cold-worked grain. And β-Nb precipitates, which have high solubility of hydrogen, retarded the precipitation of hydrides, thereby enhancing the resistance of hydride embrittlement. Fracture surface of the tensile tested specimen was observed by SEM. From the SEM observation, secondary cracks caused by hydrides were found on the fracture surface of the hydrogen-charged specimens. Cleavage facets were observed on the fracture surface of unalloyed Zr; however, a mixture of dimple and cleavage facets was observed on the surface of the Nb-containing alloys, which indicates that Nb increased the resistance to hydride embrittlement of Zr alloys.

The first step for delayed hydride cracking in zirconium alloys

Two models for delayed hydride cracking (DHC) in zirconium alloys are distinguished by their first step: – The loading of a crack induces hydride precipitation. The hydride is postulated to create a hydrogen concentration gradient, where the bulk concentration is greater than that at the crack tip. This concentration gradient is taken as the driving force for diffusion of hydrogen to the crack tip, and subsequent hydride growth. This model is called the precipitate first model (PFM). – The tensile stress at the crack tip induces a gradient in chemical potential that promotes the diffusion of hydrogen to the crack tip. Hydrides form if the hydrogen concentration reaches the solubility limit for hydride precipitation. The mechanism is postulated to create a hydrogen concentration gradient, where the bulk concentration is lower than that at the crack tip. The gradient in chemical potential is taken as the driving force for diffusion of hydrogen to the crack tip, and subsequent hydride growth. This model is called the diffusion first model (DFM). The second model, DFM, is developed. This model is shown to describe the main features of the experimental observations of DHC, without invoking new phenomena, such as reduction in the solubility limit for precipitation of hydride, as required by the PFM.

Observation of growth of a precipitate at a stress concentration by synchrotron X-ray diffraction

This paper reports X-ray diffraction results obtained during in situ precipitate growth studies. A Zr–2.5Nb specimen containing 60 wt. ppm hydrogen was subject to a thermomechanical cycle in order to precipitate hydrides preferentially at a 15 μm root radius notch. Diffraction patterns were collected as a function of time, during growth of a large (∼100 μm long) notch-tip hydride. The results indicate that hydride precipitation relaxes the crack tip strain field and confirm differences in overload behavior observed at room temperature and 250 °C.

Microstructural evolution and hydride precipitation mechanism in hydrogenated Ti–6Al–4V alloy

Use of hydrogen as a temporary alloying element in titanium alloys is an attractive approach for enhancing processability, and also for controlling the microstructure and improving final mechanical properties. In this study, the α + β titanium alloy, Ti–6Al–4V, was hydrogenated with hydrogen levels of 0.1, 0.3 and 0.5 wt%. The microstructure, phases and phase transformations were investigated by optical microscopy, X-ray diffraction and transmission electron microscopy. The results showed that the hydrogen addition had a noticeable influence on the microstructure of Ti–6Al–4V alloy. Hydrogen stabled the β-phase and leaded to the formation of hexagonal close packed α′ martensite as well as face-centered cubic δ hydride. Microstructural evolution and hydride precipitation mechanism in hydrogenated Ti–6Al–4V alloy was revealed.

Hydride Reorientation and Delayed Hydride Cracking of Spent Fuel Rods in Dry Storage

The aim of this work is to investigate the effect of thermal creep during vacuum drying of spent fuel rods on hydride reorientation and their delayed hydride cracking (DHC) susceptibility. To these ends, we analyzed Tsai’s thermal creep results of irradiated Zircaloy-4 cladding segments from two pressurized water reactors and Simpson and Ells’ observation where zirconium alloy cladding tube failed during long-term storage at room temperature. On cooling under 190 MPa, the spent fuel rods crept to 3.5 pct strain during vacuum drying showed large radial hydrides, while the ones crept to 0.35 pct strain had very fine radial hydrides. Thus, it is suggested that prior creep deformation promotes hydride reorientation in spent fuel rods on cooling after vacuum drying. Evidence for this suggestion is provided by a model experiment. Considering Kim’s DHC model and experimental facts showing precipitation of hydrides even at room temperature at stress raisers, we suggest that spent fuel rods would fail by DHC in dry storage if stress raisers are present inside the cladding on cooling to below 180 °C, and then axial splits of the failed spent fuel rods would occur by DHC due to fuel expansion by UO2 oxidation.

Dissolution and precipitation behaviors of hydrides in N18, Zry-4 and M5 alloys

The terminal solid solubilities of hydrides dissolution (TSSD) during heatup and precipitation (TSSP) during cooldown for N18, Zry-4 and M5 alloys with hydrogen concentrations of 20–240 ppm were measured by differential scanning calorimetry (DSC) in the range of 50–500 °C. There is little difference in either TSSD or TSSP solvi among these alloys, and best-fit equations were derived from the two solvi. A significant hysteresis between the solvi of TSSD and TSSP occurred, resulting from the hydride–matrix volumetric misfit strain. Based on the widths of the DSC peaks obtained during cooldown, the average precipitation rates of zirconium hydrides from the super-saturated state were evaluated by best-fit equations. The hydride precipitation is rate-controlled by hydrogen diffusion, and the average rates increased with increasing temperature.

Overload Fracture of Hydrided Region at Simulated Blunt Flaws in Zr-2.5Nb Pressure Tube Material

A crack initiation and growth mechanism known as delayed hydride cracking (DHC) is a concern for Zr-2.5Nb alloy pressure tubes of CANada Deuterium Uranium or CANDU (CANDU is a trademark of the Atomic Energy of Canada Limited, Ontario, Canada) nuclear reactors. DHC is a repetitive process that involves hydrogen diffusion, hydride precipitation, formation, and fracture of a hydrided region at a flaw tip. An overload occurs when the flaw-tip hydrided region is loaded to a stress, higher than that at which this region is formed. For the fitness-for-service assessment of the pressure tubes, it is required to demonstrate that the overload from the normal reactor operating and transient loading conditions will not fracture the hydrided region, and will not initiate DHC. In this work, several series of systematically designed, monotonically increasing load experiments are performed on specimens, prepared from an unirradiated pressure tube with hydrided region, formed at flaws with a root radius of 0.1 mm or 0.3 mm, under different hydride formation stresses and thermal histories. Crack initiation in the overload tests is detected by the acoustic emission technique. Test results indicate that the resistance to overload fracture is dependent on a variety of parameters including hydride formation stress, thermal history, hydrogen concentration, and flaw geometry.

Delayed Hydride Cracking Initiation at Notches in Zr-2.5Nb Alloys

Delayed hydride cracking (DHC) is an important crack initiation and growth mechanism in Zr-2.5Nb alloy pressure tubes of CANDU nuclear reactors. DHC is a repetitive process that involves hydrogen diffusion, hydride precipitation, growth, and fracture of a hydrided region at a flaw tip. In-service flaw evaluation requires analyses to demonstrate that DHC will not initiate from the flaw. The work presented in this paper examines DHC initiation behavior from V-notches with root radii of 15 μm, 30 μm, and 100 μm, which simulate service-induced debris fretting flaws. Groups of notched cantilever beam specimens were prepared from two unirradiated pressure tubes hydrided to a nominal hydrogen concentration of 57 wt. ppm. The specimens were loaded to different stress levels that straddled the threshold value predicted by an engineering process-zone (EPZ) model, and subjected to multiple thermal cycles representative of reactor operating conditions to form hydrides at the notch tip. Threshold conditions for DHC initiation were established for the notch geometries and thermal cycling conditions used in this program. Test results indicate that the resistance to DHC initiation is dependent on notch root radius, which is shown by optical metallography and scanning electron microscopy to have a significant effect on the distribution and morphology of the notch-tip reoriented hydrides. In addition, it is observed that one tube is less resistant to DHC initiation than the other tube, which may be attributed to the differences in their microstructure and texture. There is a reasonable agreement between the test results and the predictions from the EPZ model.

Identification and quantification of hydride phases in Zircaloy-4 cladding using synchrotron X-ray diffraction

Zirconium hydrides precipitate in fuel cladding alloys as a result of hydrogen uptake from the high-temperature corrosion environment of light water reactors. Synchrotron X-ray diffraction was performed at room temperature on stress-relieved Zircaloy-4 cladding with two distributions of hydrides – (1) uniformly distributed hydrides across the entire cladding wall and (2) hydride rim next to the outer surface. The δ-hydride phase was found to be the predominant hydride phase to precipitate for hydrogen contents up to 1250 weight parts per million (wt ppm). At a higher content, about 3000 wt ppm, although δ-hydride is still the majority phase, a significant amount of γ-hydride is also observed. At even higher hydrogen contents, in excess of approximately 6000 wt ppm, such as can occur in a highly dense hydride layer, peaks associated with the ε-hydride phase are also observed in the diffraction pattern. The volume fraction of hydrides was estimated as a function of hydrogen content using the integrated intensities of select diffraction peaks corresponding to the α-Zr matrix and the hydride phases. These estimated values agree well with calculated values from the independently measured concentrations. The results of this study indicate that hydride precipitation in Zircaloy-4 is a complex process of evolving hydride phases with increasing local hydrogen content.

Microstructure and mechanical properties of Cu–3 at.% Ti alloy aged in a hydrogen atmosphere

The microstructure of Cu–3 at.% Ti alloy aged at 773 K in a hydrogen atmosphere of 0.37 MPa has been investigated, together with hardness and tensile strength. At the early stage of aging, needle-shaped α-Cu 4Ti particles precipitate finely in a similar manner to the case of conventional vacuum-aging. On further aging, some α-Cu 4Ti precipitates are extended, while others are replaced by δ-TiH 2 precipitates. The precipitation of titanium hydride reduces the concentration of Ti in the matrix more efficiently than aging in a vacuum. The yield strength first rises, but falls on further aging. This is due to the microstructural evolution: α-Cu 4Ti particles first form and grow, but then diminish because of the formation of titanium hydrides. Elongation to fracture, as well as the electrical conductivity, is promoted by long-time aging in the hydrogen atmosphere, reflecting the reduced Ti concentration in the matrix.

Failure of high burnup fuels under reactivity-initiated accident conditions

Pulse irradiation experiments of high burnup light-water-reactor fuels were performed to assess the fuel failure limit in a postulated reactivity-initiated accident (RIA). A BWR-UO 2 rod at a burnup of 69 GW d/t failed due to pellet-cladding mechanical interaction (PCMI) in the test LS-1. The fuel enthalpy at which fuel failure occurred was comparable to those for PWR-UO 2 rods of 71 to 77 GW d/t with more corroded cladding. Comparison of cladding metallographs between the BWR and PWR fuel rods suggested that the morphology of hydride precipitation, which depends on the cladding texture, affects the fuel failure limit. The tests BZ-1 and BZ-2 with PWR-MOX rods of 48 and 59 GW d/t, respectively, also resulted in PCMI failure. The fuel enthalpies at failure were consistent with a tendency formed by the previous test results with UO 2 fuel rods, if the failure enthalpy is plotted as a function of the cladding outer oxide thickness. Therefore, the PCMI failure limit under RIA conditions depends on the cladding corrosion states including oxidation and hydride precipitation, and the same failure limit is applicable to UO 2 and MOX fuels below 59 GW d/t.

The terminal solid solubility of hydrogen in irradiated Zircaloy-2 and microscopic modeling of hydride behavior

Differential scanning calorimetry (DSC) has been applied to elucidate the terminal solid solubility (TSS) of hydrogen in Zircaloy-2 cladding tubes and spacer bands irradiated in commercial BWRs. While recovery of irradiation defects during the first heating stage of as-irradiated specimens made the DSC peak of hydride dissolution dull or broader, no significant difference was detected in the TSS between unirradiated and irradiated Zircaloy-2, irrespective of fast neutron fluence. The effect of post-irradiation annealing on TSS was also examined. The results suggest almost no interaction between irradiation defects and dissolved hydrogen or hydrides at temperatures around 300 °C. Using the present TSS data and reported hydrogen- and hydride-related properties, a microscopic analysis code HYMAC for analyzing hydride behavior in cladding tube with textured grains was constructed. Stress-induced preferential precipitation and dissolution of hydrides were reproduced by adopting a TSS sub-model in which the solubilities decrease in proportion to stress normal to the habit plane in grains and to grain faces. Analyzed results by the code were consistent with typical experimental results of hydride behavior.

Identification and characterization of a new zirconium hydride

Zirconium alloys are currently used in nuclear power plants where they are susceptible to hydrogen pick-up. Hydride precipitation may occur when the hydrogen solubility limit is reached. Various Zr hydride phases, gamma, delta and epsilon have been identified since the 1950s. Combining electron precession microdiffraction, electron energy loss spectroscopy and ab initio electronic calculations, a new Zr hydride named zeta has been identified and characterized. It belongs to the trigonal crystal system with space group P3 m1 and it is fully coherent with the alphaZr matrix.

Toward a Better Understanding of Dimensional Changes in Zircaloy-4: What is the Impact Induced by Hydrides and Oxide Layer?

Abstract In pressurized water reactors, new operating conditions (higher burnup, new chemistry, etc.) can have an effect on the dimensional stability of the fuel assembly skeleton. Previous studies have shown that the fuel assembly growth is, among others parameters (free growth, creep), strongly driven by corrosion. Oxide layer and hydrides precipitation could both induce an increase of the dimensional parameters. The scope of the present study is, regardless of irradiation effect, to quantify and to understand the separate effects of hydrogen and oxide layers on the Zircaloy-4 dimensional changes. Experimental works have been performed in laboratory on stress relieved annealed (SRA) and recrystallized (RXA) Zircaloy-4 strips. First, the hydrogen impact on dimensional changes has been studied without the effect of the oxide layer. The measurements were performed at room temperature on strips previously pre-hydrided by the gaseous charging method. The hydrogen content of the samples was between 100 ppm and 2000 ppm. Results indicate a linear correlation between hydrogen content and length variation. RXA material is more affected by the hydrogen effect than the SRA material. Nevertheless, in comparison with data issued from out-of-reactor measurement, the impact of hydrides is not sufficient (with irradiation growth) to explain the post-irradiation examinations (PIE) results. To understand these differences, the oxide layer contribution must be quantified. Second, the impact of the oxide layer was therefore studied on RXA Zircaloy-4 strips. Corrosion tests have been performed in autoclave at 360°C in primary water (2 ppm Li-1000 ppm B–H2) on as-received and pre-hydrided materials. To obtain thicker oxide layer within a shorter duration, samples have been also oxidized in furnace at 415°C. Moreover, as no significant hydriding occurs during oxidation in air, we are able to characterize properly the specific effect of the oxide layer. As for hydrides’ effect, an increase of strain is observed as the oxide thickness becomes thicker. The contributions of hydrides and oxide layer are then discussed with regard to the metallurgical properties of the alloy. Finally, all these results are compared with PIE observations. Free growth, hydride precipitation, and oxide thickness seem to be the three main parameters to explain the dimensional changes in Zircaloy-4 observed in reactor.

Evaluation of Hydride Reorientation Behavior and Mechanical Properties for High-Burnup Fuel-Cladding Tubes in Interim Dry Storage

Abstract The hydride stress reorientation behavior and the mechanical properties of irradiated cladding tubes were investigated to evaluate the high-burnup fuel-cladding tube properties in interim dry storage. As for the boiling water reactor (BWR) Zircaloy-2 (Zry-2) cladding, the hydride reorientation to the radial direction occurred at relatively low hoop stresses during the hydride reorientation treatment (HRT), such as less than 70 MPa. The increase of reorientation with hoop stress was not monotonic for the specimens in which a part of the hydrides remained precipitated at the HRT temperature, such as the case for 50GWd/t type cladding at a 300°C HRT. The degree of reorientation depended on the HRT solution temperature rather than on the estimated temperature at which the hydride precipitation occurred under the relatively moderate HRT conditions. In the relatively low cooling rate HRT, the hydride preferential precipitation in the Zr liner increased for Zr lined cladding compared to that in a relatively high cooling rate. The ductility of the specimens after the 300°C HRT showed relatively good correlation to the Polymax index which reflects the length or continuity of the hydrides regardless of their orientation. The ductility of the specimens after the 400°C, 0 MPa, 30°C/h HRT increased in ring compression testing at room temperature compared to no HRT (as-irradiated) specimens, and it indicated recovery of irradiation damage occurred at the 400°C annealing temperature and affected the ductility of the irradiated Zry-2 cladding. As for the pressurized water reactor Zircaloy-4 cladding, little increase in the radial hydride ratio occurred in a 100 MPa, 340°C or less HRT. On the other hand, the amount and the length of the hydride in the midwall area of the cladding depended on the temperature and the cooling rate from the HRT due to hydrogen migration from the hydride rim area. It is deduced that the ductility in ring compression deformation was affected by the orientation, amount, and length of hydride in the midwall area.