Hydrided Zircaloy-4 Research Articles

Determining the habit plane for radial δ-hydrides in a textured zirconium-alloy tube under hoop tensile stress: A theory-experiment approach

The orientation of hydrides with respect to externally applied stress affects the susceptibility of Zirconium (Zr) alloy fuel cladding tubes to through-wall failure. Previous studies have mainly focused on the situation in the absence of external stress, with little attention paid to the situation under applied stress. Hoop tensile stress can cause the reorientation of hydride platelets from the circumferential to the radial direction in the radial textured Zr fuel clad tubes, substantially degrading its hoop ductility. In this study, we used a theory-experiment approach to investigate the radial hydride-matrix orientation relationship. A microstructure criterion was proposed to determine the habit plane of the radial δ-hydrides. Our results demonstrate that radial δ-hydrides predominantly form on the pyramidal 101¯1α plane of α-Zr with 101¯1α−Zr//111δ−hydride. The electron backscatter diffraction study of hydrided Zircaloy-4 tube provides experimental evidence supporting theoretical prediction. The underlying mechanism for this phenomenon is attributed to a competition between tensile stress-assisted hydride nucleation and strain-induced resistance to nucleation.

Revealing the role of hydrogen content on the mechanical behavior and cracking mechanisms of Zircaloy-4 under strain

The mechanical property of zirconium-alloy reactor cladding materials inevitably degrades because of hydrogen absorption and hydride precipitation. In this study, mechanical behavior and delayed hydrogen cracking (DHC) mechanisms of Zircaloy-4 alloys with a broad range of hydrogen contents were investigated. Scanning electron microscopy (SEM), electron backscatter diffraction (EBSD), transmission electron microscopy (TEM), and a universal testing machine were jointly employed for this purpose. Results showed that the mechanical property of the hydrogenated specimen not only depended on the size and orientation of hydride, but also was affected by the continuity degree of the load-bearing matrix. The relevant microstructural observation indicated that the lower continuity degree of the load-bearing matrix, the less ductility of Zircaloy-4. More importantly, hydride phase transformation from δ-ZrH 1.5 to γ-ZrH occurred around the crack tip and reduced the local strain, while twinning just appeared at the crack bridging sites in the higher hydrogen-content specimens for changing the crystal orientation and mitigating the local stress concentration. Both those effects are expected to be potential mechanisms of the reappearing yield phenomenon because they are closely linked to dislocation movement. Additionally, based on the combination effect of crack blunting of γ-ZrH and the strain accommodation of twinning, the trend of crack propagation is delayed locally as expected. • The hydride phase transformation and twinning phenomenon around cracks were analyzed using EBSD. • The continuity of the load-bearing matrix profoundly influenced the mechanical properties of hydrided zircaloy-4. • The appearance and disappearance of yield-suppression phenomenon in hydrided zircaloy-4 were investigated. • Both γ-ZrH and twins are beneficial to delay the localized crack propagation.

Determination of the hydrogen heat of transport in Zircaloy-4

During operation in a nuclear reactor, Zr-based nuclear fuel cladding is subject to waterside corrosion which can lead to hydrogen ingress. The hydrogen that enters the material will migrate to colder spots and precipitate as zirconium hydrides if the hydrogen content exceeds the hydrogen terminal solid solubility in the material. Since a temperature gradient is established in the radial direction of the cladding during operation, the hydrides can preferentially precipitate at the colder outer surface of the cladding. Other gradients can also occur in the longitudinal and azimuthal directions of the cladding tube. As a consequence, hydrogen redistributes itself in response to the concentration and temperature gradients present in the sample. The response of the hydrogen in solid solution to temperature gradients is governed by the heat of transport Q* as a function of temperature, so it can be used in the BISON code. A set of experiments was set up to determine the heat of transport (Q*), in which a uniformly hydrided Zircaloy-4 sample is annealed under a fixed temperature gradient at a range of temperatures, and the resulting hydrogen distribution is analyzed to determine Q*. The results are discussed in terms of existing literature.

Ductile tearing criteria and failure probability estimation of hydrided Zircaloy-4 cladding under axial loads

Ductile tearing criteria and failure probability estimation of hydrided Zircaloy-4 cladding under axial loads

Oxidation Behavior Characterization of Zircaloy-4 Cladding with Different Hydrogen Concentrations at 500-800 °C in an Ambient Atmosphere.

This study simulated the after-burned zirconium cladding oxidation in air at temperatures between 500 and 800 °C. The weight changes of Zircaloy-4 cladding with hydrogen contents of 100–1000 ppm continuously measured through thermogravimetric analysis (TGA) during oxidation tests at different temperatures in an air atmosphere. The TGA results indicate a transition of oxidation kinetics from a parabolic rate law to a linear rate law for as-received and hydrided Zircaloy-4 cladding. The hydrogen concentration of Zircaloy-4 had a marked effect on its pre-transition oxidation in air between 500 and 800 °C. For all samples, the linear oxidation (post transition stage) at 650 °C, which is the critical oxidation temperature, displays a similar trend. In addition, scanning electron microscopy and transmission electron micros-copy examinations indicated the presence of a few and numerous discontinuous micro-cracks in the oxide layer in the pre-transition and post-transition stages, respectively.

Hydride embrittlement resistance of Zircaloy-4 and Zr-Nb alloy cladding tubes and its implications on spent fuel management

• Mechanical tests of hydrided Zircaloy-4 and Zr-Nb alloy with Ring Compression Test. • EBSD analyses on texture of reactor-grade hydrided Zircaloy cladding tubes. • Abrupt ductile to brittle transition at 560 and 490 wppm for Zr-4 and Zr-Nb alloy. • Larger grain size of Zr-4 reduces interlink of inter-granular hydrides. • Reduced hydride interlink increases hydride embrittlement resistance. In the spent fuel storage phase, nuclear fuel cladding is subjected to increased embrittlement owing to a large amount of hydride precipitation. This study compares the differences in the hydrogen-induced cladding embrittlement of cold work stress-relief annealed (CWSR or SRA) Zircaloy–4 and Zr-Nb alloy cladding with ring compression test at the temperature of the spent fuel pool, which is approximately 40 °C. Experiments demonstrate that an abrupt ductile to brittle (DTB) transition occurs at the critical hydrogen content of 560 and 490 wppm for Zircaloy-4 and the tested Zr-Nb cladding tubes, respectively. Even beyond the critical hydrogen content, sufficiently high cladding ductility with the offset strain >10% is maintained up to ∼ 90 MWd/kgU for both cladding materials on the rod-average basis. Extensive EBSD analyses coupled to thermodynamic modeling demonstrate that this is primarily due to the slightly larger grain diameter of Zircaloy-4 tube, which reduces the number of available sites for inter-granular hydride precipitation. Reduced inter-granular hydride precipitation prevents the extent of hydride interlink, thereby improving the hydride embrittlement resistance. Nevertheless, the tested Zr-Nb alloy cladding presents an extened discharge burnup limit for abrupt DTB transition owing the reduced in-core cladding oxidation rate. The presented understanding of microstructural effect on hydride interlink and resulting embrittlement may provide a basis for understanding the general hydride embrittlement phenomena of Zircaloy cladding which include, but not limited to, wet storage, dry storage, and post-accident ductility of high burnup Zircaloy cladding.

Temperature-dependent axial mechanical properties of Zircaloy-4 with various hydrogen amounts and hydride orientations

The effects of hydride amount (20–850 wppm), orientation (circumferential and radial), and temperature (room temperature, 100 °C, 200 °C) on the axial mechanical properties of Zircaloy-4 cladding were comprehensively examined. The fraction of radial hydride fraction in the cladding was quantified using PROPHET, an in-house radial hydride fraction analysis code. Uniaxial tensile tests (UTTs) were conducted at various temperatures to obtain the axial mechanical properties. Hydride orientation has a limited effect on the axial mechanical behavior of hydrided Zircaloy-4 cladding. Ultimate tensile stress (UTS) and associated uniform elongation demonstrated limited sensitivity to hydride content under UTT. Statistical uncertainty of UTS was found small, supporting the deterministic approach for the load-failure analysis of hydrided Zircaloy-4 cladding. These properties notably decrease with increasing temperature in the tested range. The dependence of yield strength on hydrogen content differed from temperature to temperature. The ductility-related parameters, such as total elongation, strain energy density (SED), and offset strain decrease with increasing hydride contents. The abrupt loss of ductility in UTT was found at ∼700 wppm. Demonstrating a strong correlation between total elongation and offset strain, SED can be used as a comprehensive measure of ductility of hydrided zirconium alloy.

Determining failure properties of as-received and hydrided unirradiated Zircaloy-4 from ring compression tests

An integrated experimental and computational approach was developed to determine the plastic strain at fracture and to investigate the fracture energy of as-received and hydrided unirradiated Zircaloy-4 nuclear cladding tubes in the hoop direction at room temperature from ring compression testing (RCT). This work builds on previous methods in the literature that were developed to obtain the mechanical properties of nuclear cladding before fracture (Young’s modulus, yield stress, and strain-hardening parameters) from experimental RCT results and extends the characterization of this material class to include material failure properties. A ductile damage approach was developed and implemented in a validated finite element model to predict material failure, evaluate fracture energy, and quantify the plastic strain at fracture initiation. The hydrogen content of the Zircaloy-4 specimens was varied up to 610 wppm when the failure became brittle. As expected, increasing hydrogen content caused embrittlement of the Zircaloy-4 samples tested and the plastic strain at fracture and fracture energy to decrease

Fracture-mechanics-based evaluation of failure limit on pre-cracked and hydrided Zircaloy-4 cladding tube under biaxial stress states

ABSTRACTTo better understand the failure limit of fuel cladding during the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA), pre-cracked and hydrided cladding samples with base metal final heat-treatment status of cold worked (CW) and recrystallized (RX) were tested under biaxial stress conditions (strain ratios (εz/εθ) of 0 and 0.5). Displacement-controlled biaxial-expansion-due-to-compression (biaxial-EDC) tests were performed to obtain the hoop strain at failure (failure strain) of the samples. The conversion of the failure strains to J-integrals at failure (JIC) by finite-element analysis involving data of stress-relieved (SR) cladding specimens from our previous study revealed that the failure limit in the dimension of JIC unifies the effects of pre-crack depth. About 30–50% reduction in the JIC value was observed as the strain ratio increased from 0 to 0.5 irrespective of the annealing type, pre-crack depth, and hydrogen content. The rate of fractional decreases of JIC with increase of hydrogen content is in the order of CW>SR>RX, which are essentially independent of strain ratio for the CW and SR samples. The results were incorporated into the failure prediction model of the JAEA’s fuel performance code in the form of a correction factor that considers the biaxial loading effect.

Effects of hydride rim on the ductility of Zircaloy-4 cladding

High burn-up nuclear fuel cladding has a hydride rim owing to hydrogen diffusion under a temperature gradient. To simulate the high burn-up cladding, the hydride rim was formed in un-irradiated Zircaloy-4 cladding and ductility evaluation was conducted using ring compression test (RCT). The RCTs were conducted from room temperature to 300 °C for determining the ductile to brittle transition temperature (DBTT) of hydrided Zircaloy-4 cladding. Hydride rimmed specimens have lower ductility and DBTT than uniform hydrided specimens. The increased ductility at high temperatures is likely due to the increased ductility of the Zr matrix because the fractography at hydride rim still shows brittle fracture at those temperatures. The ductility of hydride rimmed specimens with radial hydrides are compared with those of irradiated Zr alloy claddings, and thus, it is concluded that hydride rimmed specimen can simulate the high burn-up Zr alloy cladding to some extent.

Application of configurational mechanics in characterizing contact fatigue life and its crack propagation: a numerical lattice-based approach

A numerical lattice-based approach equipped with a brittle erosion algorithm and material forces in the context of LEFM and configurational mechanics is introduced to analyze contact fatigue life and crack propagation of brittle solids under low amplitude cyclic loading using the classical Paris’s law. Material force vectors at lattice nodes were employed in identifying potential crack path in high-cycle fatigue applications. A 2D plane strain pad-substrate system with contact formulations under constant compression and cyclic sinusoidal shear with constant amplitude was considered to investigate fatigue crack’s growth rate and its shape in the substrate, made of hydrided brittle Zircaloy-4 used in nuclear industry. The capability of the lattice in predicting the crack extension path using the material force vectors at the crack tip is verified by comparing with analytical solutions for a centered crack domain under pure shear and direct tension loading, and also for an initial small surface crack for the contact fatigue pad-substrate system. Obtaining the total fatigue life by assuming a failure value for the crack length and confirming the Paris’ law crack growth, the lattice results demonstrate that the fatigue crack path in the substrate is a curved trajectory which gradually reduces as the fatigue crack tip grows deeper into the substrate. Having simple constitutive formulation and straightforward erosion algorithm, this approach together with configurational mechanics implications can be employed in analyzing fretting fatigue problems.

Biaxial creep performance of CWSR Zircaloy-4 cladding at emulated off-normal conditions of interim dry storage facility

Biaxial creep behavior of hydrided Zircaloy-4 cladding was tested at 500 °C. Creep-life was inversely proportional to H-concentration up to 750 wppm. The weakening through hydride reorientation of hydrided specimens was not observed by optical microscopy. The power-law stress exponents and the TEM observation supported the high-temperature climb mechanism dominated the creep behavior in secondary stage.

Characterization of local hydride re-orientation in high burn-up PWR fuel rods induced by high pressure at high temperatures

ABSTRACTHydride re-orientation in high burn-up PWR fuel rods was induced by high pressure at high temperatures. The high-burnup specimens were sectioned from PWR rods taken from a 15×15 assembly of the H.B. Robinson (HBR) Unit 2 reactor. Out-of cell benchmark tests performed on unirradiated hydrided Zircaloy-4 specimens were conducted to determine the appropriate temperature, pressure, cooling rate, and number of cooling cycles for the reorientation of the irradiated in-cell specimens. The in-cell hydride reorientation tests were performed using highburnup fuel specimens under a hoop stress ≈145 MPa at 400ºC. The specimens were heated to the target temperature of 400ºC, held for 3 hours, cooled at 1ºC/min to 170ºC,and then heated at 1ºC/min to the target temperature again for five cycles. Post test metallographic examinations showed that a significant amount of radial hydrides were induced in the HBR fuel rods. The length of radial hydride was up to 60 µm. For unirradiated materials, the ductility of the radial hydride treated specimens is significantly reduced as compared to the as-hydrided specimens having the same hydrogen concentration (≈300 wppm in this work). The mechanical testing on irradiated fueled samples with hydride reorientation experiments have been performed, and will be reported separately in the near future.

Hydride reorientation in Zircaloy-4 examined by in situ synchrotron X-ray diffraction

The phenomenon of stress-reorientation has been investigated using in situ X-ray diffraction during the thermomechanical cycling of hydrided Zircaloy-4 tensile specimens. Results have shown that loading along a sample’s transverse direction (TD) leads to a greater degree of hydride reorientation when compared to rolling direction (RD)-aligned samples. The elastic lattice micro-strains associated with radially oriented hydrides have been revealed to be greater than those oriented circumferentially, a consequence of strain accommodation. Evidence of hydride redistribution after cycling, to α-Zr grains oriented in a more favourable orientation when under an applied stress, has also been observed and its behaviour has been found to be highly dependent on the loading axis. Finally, thermomechanical loading across multiple cycles has been shown to reduce the difference in terminal solid solubility of hydrogen during dissolution (TSSD,H) and precipitation (TSSP,H).

The influence of stress state on the reorientation of hydrides in a zirconium alloy

Hydride reorientation can occur in spent nuclear fuel cladding when subjected to a tensile hoop stress above a threshold value during cooling. Because in these circumstances the cladding is under a multiaxial stress state, the effect of stress biaxiality on the threshold stress for hydride reorientation is investigated using hydrided CWSR Zircaloy-4 sheet specimens containing ∼180 wt ppm of hydrogen and subjected to a two-cycle thermo-mechanical treatment. The study is based on especially designed specimens within which the stress biaxiality ratios range from uniaxial (σ2/σ1 = 0) to “near-equibiaxial” tension (σ2/σ1 = 0.8). The threshold stress is determined by mapping finite element calculations of the principal stresses and of the stress biaxiality ratio onto the hydride microstructure obtained after the thermo-mechanical treatment. The results show that the threshold stress (maximum principal stress) decreases from 155 to 75 MPa as the stress biaxiality increases from uniaxial to “near-equibiaxial” tension.

A model to describe the mechanical behavior and the ductile failure of hydrided Zircaloy-4 fuel claddings between 25 °C and 480 °C

A model is proposed to describe the mechanical behavior and the ductile failure at 25, 350 and 480 °C of Zircaloy-4 cladding tubes, as-received and hydrided up to 1200 wt. ppm (circumferential hydrides). The model is based on the Gurson–Tvergaard–Needleman model extended to account for plastic anisotropy and viscoplasticity. The model considers damage nucleation by both hydride cracking and debonding of the interface between the Laves phase precipitates and the matrix. The damage nucleation rate due to hydride cracking is directly deduced from quantitative microstructural observations. The other model parameters are identified from several experimental tests. Finite element simulations of axial tension, hoop tension, expansion due to compression and hoop plane strain tension experiments are performed to assess the model prediction capability. The calibrated model satisfactorily reproduces the effects of hydrogen and temperature on both the viscoplastic and the failure properties of the material. The results suggest that damage is anisotropic and influenced by the stress state for the non-hydrided or moderately hydrided material and becomes more isotropic for high hydrogen contents.

In situ micropillar deformation of hydrides in Zircaloy-4

Deformation of hydrided Zircaloy-4 has been examined using in situ loading of hydrided micropillars in the scanning electron microscope and using synchrotron X-ray Laue microbeam diffraction. Results suggest that both the matrix and hydride can co-deform, with storage of deformation defects observed within the hydrides, which were twinned. Hydrides placed at the plane of maximum shear stress showed deformation within the hydride packet, whilst packets in other pillars arrested the propagation of shear bands. X-ray Laue peak broadening, prior to deformation, was associated with the precipitation of hydrides, and during deformation plastic rotation and broadening of both the matrix and hydride peaks were observed. Post-mortem TEM of the deformed pillars has indicated a greater density of dislocations associated with the precipitated hydride packets, while the observed broadening of the hydride electron diffraction spots further suggests that plastic strain gradients were induced in the hydrides by compression.

Microstructure and texture analysis of δ-hydride precipitation in Zircaloy-4 materials by electron microscopy and neutron diffraction

This work presents a detailed microstructure and texture study of various hydrided Zircaloy-4 materials by neutron diffraction and microscopy. The results show that the precipitated δ-ZrH1.66generally follows the δ(111)//α(0001) and δ[1{\overline 1}0]//α[11{\overline 2}0] orientation relationship with the α-Zr matrix. The δ-hydride displays a weak texture that is determined by the texture of the α-Zr matrix, and this dependence essentially originates from the observed orientation correlation between α-Zr and δ-hydride. Neutron diffraction line profile analysis and high-resolution transmission electron microscopy observations reveal a significant number of dislocations present in the δ-hydride, with an estimated average density one order of magnitude higher than that in the α-Zr matrix, which contributes to the accommodation of the substantial misfit strains associated with hydride precipitation in the α-Zr matrix. The present observations provide an insight into the behaviour of δ-hydride precipitation in zirconium alloys and may help with understanding the induced embrittling effect of hydrides.

The influence of hydride on fracture toughness of recrystallized Zircaloy-4 cladding

In this work, RXA cladding tubes were hydrogen-charged to target hydrogen content levels between 150 and 800wppm (part per million by weight). The strings of zirconium hydrides observed in the cross sections are mostly oriented in the circumferential direction. The fracture toughness of hydrided RXA Zircaloy-4 cladding was measured to evaluate its hydride embrittlement susceptibility. With increasing hydrogen content, the fracture toughness of hydrided RXA cladding decreases at both 25°C and 300°C. Moreover, highly localized hydrides (forming a hydride rim) aggravate the degradation of the fracture properties of RXA Zircaloy-4 cladding at both 25°C and 300°C. Brittle features in the form of quasi-cleavages and secondary cracks were observed on the fracture surface of the hydride rim, even for RXA cladding tested at 300°C.

Hydride precipitation and its influence on mechanical properties of notched and unnotched Zircaloy-4 plates

The hydride formation and its influence on the mechanical performance of hydrided Zircaloy-4 plates containing different hydrogen contents were studied at room temperature. For the unnotched plate samples with the hydrogen contents ranging from 25 to 850wt.ppm, the hydrides exerted an insignificant effect on the tensile strength, while the ductility was severely degraded with increasing hydrogen content. The fracture mode and degree of embrittlement were strongly related to the hydrogen content. When the hydrogen content reached a level of 850wt.ppm, the plate exhibited negligible ductility, resulting in almost completely brittle behavior. For the hydrided notched plate, the tensile stress concentration associated with the notch tip facilitated the hydride accumulation at the region near the notch tip and the premature crack propagation through the hydride fracture during hydriding. The final brittle through-thickness failure for this notched sample was mainly attributed to the formation of a continuous hydride network on the thickness section and the obtained very high hydrogen concentration (estimated to be 1965wt.ppm).