Stress-reorientation of hydrides and hydride embrittlement of Zr-2.5 wt% Nb pressure tube alloy

Hydrogen in excess of the terminal solid solubility precipitates out as a brittle hydride phase in zirconium alloys. The hydrides acquire platelet shaped morphology due to their accommodation in the matrix and can cause severe embrittlement, especially when these are oriented normal to the tensile stress axis. The precipitation of hydride platelets normal to the tensile stress when cooled under stress from a solution-annealing temperature is commonly referred to as ‘stress-reorientation’. The stress-reorientation is associated with a threshold stress below which no reorientation is observed. In this work, stress-reorientation of hydrides was investigated for unirradiated, cold worked and stress-relieved Zr-2.5 wt% Nb pressure tube material for a reorientation temperature of 423-723 K. The effect of the reoriented hydrides on the tensile properties of the Zr-2.5 wt% Nb pressure tube alloy was evaluated in the temperature range of 298-573 K.     

The influence of hydride blisters on the fracture of Zircaloy-4

The fracture behavior under near plane-strain deformation conditions of Zircaloy-4 sheet containing solid hydride blisters of various depths has been examined at 25 and 300 °C. The study was based on material with either model ‘blisters’ having diameters of 2 and 3 mm or a continuous layer of hydride; in all cases, the substrate material contained discrete hydride precipitates. The fracture strains decrease rapidly with increasing hydride blister/layer depth to levels of about 100 μm deep, and then remain roughly constant. For a given blister depth, the material is significantly more ductile at 300 °C than at room temperature although measurable ductility is retained even at 25 °C and for large blister depths. The material is somewhat more ductile if the hydride is in the form of a blister than in the form of a continuous layer (rim). The hydride blisters/layers are brittle at all temperatures, and crack shortly after yielding of the ductile substrate. Consequently, both experimental evidence and analytical modeling indicate that fracture of the sheet is controlled by the crack growth resistance of the substrate at 25 °C. At elevated temperatures, the hydride particles within the substrate are quite ductile, inhibit crack growth, and failure eventually occurs due to a shear instability.     

Volume conservation during irradiation growth of Zr-2.5Nb

Dimensional changes are reported in three dimensions for cold-worked Zr-2.5 Nb pressure tube material irradiated to a fast fluence of 174 × 10²⁴ nm⁻², E > 1 MeV at a nominal temperature of 250 °C. The observed dimensional changes in the longitudinal and transverse directions (up to ∼1.2% and ∼−0.5%, respectively) are consistent with earlier data at 280 °C and 310 °C, and the previously reported negative temperature dependence. The observed growth in the radial direction is negative (up to ∼0.7%). Initially, there is a small volume increase (0.05-0.1%) but this gradually decays to < 0.05% and the long term rate of volume change is negligible, within the accuracy of the measurement, demonstrating that the phenomenon observed is, indeed, irradiation growth.     

Micro-texture of extruded Zr-2.5Nb tubes

We report the micro-texture of two extruded Zr-2.5Nb tubes determined using scanning electron microscopy combined with electron back-scattering diffraction (SEM/EBSD) and transmission electron microscopy and selected area diffraction (TEM/SAD). The pole figures determined by SEM/EBSD correspond well with bulk pole figures previously determined by X-ray diffraction (XRD). Three components of texture are seen to correlate with the shape and morphology of the α-grains and their contained dislocation substructures. The first component corresponds to elongated alpha grains containing a high density of a and c + a dislocations in which the c-axis is oriented at a relatively high angle to the long dimension of the α-grains as viewed in transverse section; these grains comprise a texture component with the c-axes in the radial transverse plane, tilted towards the radial direction. The second component corresponds to elongated α-grains containing a low dislocation density in which the c-axis is oriented parallel to the long dimension of the alpha grains: these grains also comprise a texture component with the c-axis in the radial/transverse plane, but predominantly in the transverse direction. The final component corresponds to colonies of Widmanstätten-like α-grains that are transformed from the β-phase: the majority of these grains have their c-axes in the axial direction. These grain have very low dislocation densities and are probably developed during cooling, after extrusion.     

Evaluation of effect of hydrogen on toughness of Zircaloy-2 by instrumented drop weight impact testing

Hydride-assisted degradation in fracture toughness of Zircaloy-2 was evaluated by carrying out instrumented drop-weight tests on curved Charpy specimens fabricated from virgin pressure tube. Samples were charged to 60 ppm and 225 ppm hydrogen. Ductile-to-brittle-transition behaviour was exhibited by as-received and hydrided samples. The onset of ductile-to-brittle-transition was at about 130 °C for hydrided samples, irrespective of their hydrogen content. Dynamic fracture toughness (K_ID) was estimated based on linear elastic fracture mechanics (LEFM) approach. For fractures occurring after general yielding, the fracture toughness was derived based on equivalent energy criterion. Results are supplemented with fractography. This simple procedure of impact testing appears to be promising for monitoring service-induced degradation in fracture toughness of pressure tubes.     

Dissolution and precipitation behavior of hydrides in Zircaloy-2 and high Fe Zircaloy

In order to elucidate the terminal solid solubility during the dissolution of hydrides at heatup (TSSD) and during the precipitation of hydrides at cooldown (TSSP) for hydrogenated Zircaloy-2 and high Fe Zircaloy, differential scanning calorimetry (DSC) measurements have been carried out in the temperature range of 50-600 °C. The hydrogen concentrations in the two kinds of alloys ranged from 40 to 542 ppm. There was no difference in either TSSD or TSSP solvi between Zircaloy-2 and high Fe Zircaloy, and best-fit equations were derived for the two curves. In the present TSSP data, two different activation energies, separating into high and low temperature ranges at 260 °C, were obtained. Based on the widths of the DSC peak obtained during cooldown, the average precipitation (nucleation plus growth) rates of zirconium hydrides from super-saturated state were assessed. The activation energy of the precipitation rate was approximately equivalent to reported values of hydrogen diffusion coefficients of Zr and Zircaloys.     

Influence of temperature on threshold stress for reorientation of hydrides and residual stress variation across thickness of Zr-2.5Nb alloy pressure tube

Threshold stress, σ_th, for reorientation of hydrides in cold worked and stress-relieved (CWSR) Zr-2.5Nb pressure tube material was determined in the temperature range of 523-673 K. Using tapered gage tensile specimen, mean value of σ_th was experimentally determined by two methods, half thickness method and area compensation method. The difference between local values of σ_th measured across the thickness of the tube and the mean σ_th values yielded the residual stress variation across the tube thickness. It was observed that both the mean threshold stress and residual stress decrease with increase in reorientation temperature. Also, the maximum value of residual stresses was observed near the midsection of the tube.     

Heat capacity of hydrogenated Zircaloy-2 and high Fe Zircaloy

Heat capacities (C_p) of non-hydrogenated and hydrogenated Zircaloy-2 and high Fe Zircaloy were measured in the temperature range from 350 to 873 K, using a differential scanning calorimeter. The hydrogen concentrations in the two types of alloys ranged from 26 to 1004 ppm. The C_p values of the as-received alloys with 26-29 ppm hydrogen were in good agreement with literature data for low hydrogen Zircaloys. From this finding and observation of almost the same enthalpy changes for hydride dissolution for both alloys, it was concluded that there was no difference in C_p values between the two types of hydrogenated Zircaloys. The dissolution enthalpy of hydrides calculated from C_p data was 41.0 kJ/g-atom H. For Zircaloy-2 samples with higher hydrogen concentrations than 700 ppm, the phase transition from α+δ to α+β was observed at the eutectoid temperature of 824-827 K. Two types of models describing an additional heat capacity due to the hydride dissolution were presented based on the present C_p data and previously derived terminal solid solubility of hydrogen.     

Textural and microstructural developments during fabrication of Zr-2.5Nb pressure tubes

Zr-2.5wt%Nb pressure tubes, as used in the Indian pressurized heavy water reactors (PHWR), are fabricated through a combination of hot extrusion followed by two stages of cold pilgering and annealing. The present study makes an effort to systematically characterize the textural changes during the fabrication stages. The starting single-phase hcp martensitic structure was textured to start with and it also went through strong texture developments during hot extrusion. First and especially the second pilgering modified the texture. Such modifications were related to local discontinuity of the softer second phase, as an apparent continuity restricted lattice rotation in the primary hcp phase. Annealing caused discontinuity or spherodization of the bcc phase, but did not create recrystallization in the hcp matrix. The combination of two pilgering plus annealing operations, though the latter did not cause noticeable texture changes, however, reverted the final texture close to the parent hot-extruded texture.     

The determination of hydrogen and deuterium in Zr-2.5Nb material by hot vacuum extraction mass spectrometry

An analysis method for the determination of H and D concentrations in Zr-2.5Nb material has been established based on hot vacuum extraction and isotope dilution mass spectrometry. Hot vacuum extraction enables complete removal of the hydrogen isotopes from the sample. The isotope dilution technique is used to determine quantitatively the amount of hydrogen isotopes in the extracted gas. Methods for preparing standards of H, D or H and D in Zr-2.5Nb have also been established. These ‘in-house’ standards are used to assess the performance of the analysis method. The analysis uncertainty, based on the determined content relative to the dosed H or D isotope content of each standard, is 1% (at a level of confidence of 95%) for samples containing greater than ∼5 μmoles H or D. The uncertainty increases to 5% as the sample content decreases to 0.5 μmoles H or D. The uncertainty of this analysis method is well within the requirements for surveillance examinations of CANDU^® reactors and post-irradiation examinations of reactor components.     

Influence of metallurgical variables on delayed hydride cracking in Zr-Nb pressure tubes

During service, Zr-2.5Nb pressure tubes of nuclear power reactors may be prone to suffer from crack growth by delayed hydride cracking (DHC). For a given hydrogen plus deuterium concentration there is a critical temperature (T_C) below which DHC may occur. In this work, T_C was measured for specimens cut from pressure tubes made in Canada (CANDU) and in Russia (RBMK). Hydrogen was added to the specimens to get concentrations ranging from 24 to 60 wt ppm. It was found that T_C was higher than the corresponding precipitation temperature. The crack propagation velocity (V_P), measured in axial direction, increases from a minimum at T_C to a maximum at a temperature close but higher than the precipitation temperature. At lower temperatures, when hydride precipitates are present in the bulk, V_P follows an Arrhenius law: V_P = A exp(−Q/RT), with an activation energy Q of 66-68 kJ/mol for both tubes. The RBMK material presented lower velocities than CANDU one.     

Mechanisms of stress relief cracking in titanium stabilised austenitic stainless steel

The heat affected zone (HAZ) of AISI 321 welds may exhibit a serious form of cracking during service at high temperature. This form of damage, called ‘stress relief cracking’, is known to be due to work hardening but also to aging due to Ti(C,N) precipitation on dislocations which modifies the mechanical behaviour of the HAZ. The present study aims to analyse the latter embrittlement mechanism in one specific heat of 321 stainless steel. To this end, different HAZs are simulated using an annealing heat-treatment, followed by various cold rolling and aging conditions. Then, we study the effects of work hardening and aging on Ti(C,N) precipitation, on the mechanical (hardness, tensile and creep) behaviour of the simulated HAZs and on their sensitivity to intergranular crack propagation through stress relaxation tests performed on pre-cracked CT type specimens tested at 600 °C. It is shown that work hardening is the main parameter of the involved mechanism but that aging does not promote crack initiation although it leads to titanium carbide precipitation. Therefore, the role of Ti(C,N) precipitation on stress relief cracking mechanisms is discussed. An attempt is made to show that solute drag effects are mainly responsible for this form of intergranular damage, rather than Ti(C,N) precipitation.     

Characterization of hydrogen concentration in Zircaloy-4 using ultrasonic techniques

The relationship between hydrogen concentration precipitated as hydride particles and ultrasonic parameters, such as velocity and attenuation, was examined in Zircaloy-4 samples for potential applications in the Non-Destructive Test Field. Different amounts of hydrogen (up to 517 ppm) were introduced in the samples by gaseous charging. Ultrasonic attenuation measurements were performed with compressive waves at frequencies of 10 and 30 MHz, and propagation velocity measurements were performed at 10 MHz. Ultrasonic velocity showed an approximately linear increase with hydrogen concentration and it could be used as an assessment parameter when the hydrogen level is high enough. Attenuation versus hydrogen concentration has been fitted by a logarithmic equation at 10 MHz. At 30 MHz a fluctuating behavior of the attenuation prevented measurement of the hydrogen concentration.     

Effect of radial hydrides on the axial and hoop mechanical properties of Zircaloy-4 cladding

The effect of radial hydrides on the mechanical properties of stress-relief annealed Zircaloy-4 cladding was studied. Specimens were firstly hydrided to different target hydrogen levels between 100 and 600 wt ppm and then thermally cycled in an autoclave under a constant hoop stress to form radial hydrides by a hydride reorientation process. The effect of radial hydrides on the axial properties of the cladding was insignificant. On the other hand, the cladding ductility measurements decreased as its radial hydride content increased when the specimen was tested in plane strain tension. A reference hydrogen concentration for radial hydrides in the cladding was defined for assessing the fuel cladding integrity based on a criterion of the tensile strength 600 MPa. The reference hydrogen concentration increased with the specimen (bulk) hydrogen concentration to a maximum of ∼90 wt ppm at the bulk concentration ∼300 wt ppm H and then decreased towards higher concentrations.     

Radionuclide release from high burnup spent fuel during corrosion in salt brine in the presence of hydrogen overpressure

In the case of a contact between groundwater and Fe-based spent fuel disposal containers in a repository large amounts of hydrogen will be produced by the corrosion of iron, which may result in significant hydrogen pressures. To quantify to what extent the hydrogen overpressure may counteract radiolysis enhanced matrix dissolution, related experimental work has been performed. High burnup spent fuel was corroded in 5.6 mol (kg H₂O)⁻¹ NaCl solution applying H₂ overpressures (experimental set 1) <0.17 bar by radiolysis, (experimental set 2) 2.8 bar by Fe corrosion, (experimental set 3) 3.2 bar by external H₂ gas. In the absence of Fe (experimental set 3) the UO₂ matrix dissolution rate decreased by a factor of about 10. In this test the concentrations of U, Np, Tc in solution were found to be decreasing by at least two orders of magnitude, and ranging within the same level as in the presence of Fe powder (experimental set 2). However, Pu and Am concentrations (experimental set 3) were less affected, due to the high sorption capacity for these radioelements onto Fe corrosion products.     

Micro-engineered first wall tungsten armor for high average power laser fusion energy systems

The high average power laser program is developing an inertial fusion energy demonstration power reactor with a solid first wall chamber. The first wall (FW) will be subject to high energy density radiation and high doses of high energy helium implantation. Tungsten has been identified as the candidate material for a FW armor. The fundamental concern is long term thermo-mechanical survivability of the armor against the effects of high temperature pulsed operation and exfoliation due to the retention of implanted helium. Even if a solid tungsten armor coating would survive the high temperature cyclic operation with minimal failure, the high helium implantation and retention would result in unacceptable material loss rates. Micro-engineered materials, such as castellated structures, plasma sprayed nano-porous coatings and refractory foams are suggested as a first wall armor material to address these fundamental concerns. A micro-engineered FW armor would have to be designed with specific geometric features that tolerate high cyclic heating loads and recycle most of the implanted helium without any significant failure. Micro-engineered materials are briefly reviewed. In particular, plasma-sprayed nano-porous tungsten and tungsten foams are assessed for their potential to accommodate inertial fusion specific loads. Tests show that nano-porous plasma spray coatings can be manufactured with high permeability to helium gas, while retaining relatively high thermal conductivities. Tungsten foams where shown to be able to overcome thermo-mechanical loads by cell rotation and deformation. Helium implantation tests have shown, that pulsed implantation and heating releases significant levels of implanted helium. Helium implantation and release from tungsten was modeled using an expanded kinetic rate theory, to include the effects of pulsed implantations and thermal cycles. Although, significant challenges remain micro-engineered materials are shown to constitute potential candidate FW armor materials.     

Damage inhomogeneity in the core region of displacement cascades in simplified nuclear glasses

Displacement cascades at energies ranging from 16 keV to 70 keV were simulated by classical molecular dynamics. Damage inhomogeneity was observed in each case: the atomic density was diminished by the incident projectile to a variable extent depending on the regions concerned. The regions near the initial projectile position are largely annealed, and regions near the end of the cascade are relatively unaffected because of the low residual projectile energy. However, maximum damage occurs in intermediate regions from collisions with incident projectiles at energies ranging from about 10 keV to 25 keV. This phenomenon illustrates the competition between structure annealing and projectile-induced damage: both increase with the local energy, but with different dynamics. At the highest energies, annealing wins out over damage, restoring the glass structure to its pristine state; hence the good structural behaviour in the zones closest to the initial projectile position, which are subjected to the greatest local temperature rise.     

High-pressure hydriding of Zircaloy cladding by the thermogravimetry and tube-burst techniques

Hydriding kinetics of modified Zircaloy claddings was studied by the thermogravimetric method at 400 °C and the tube-burst technique at 315 °C. Some specimens were prefilmed with a thin oxide layer by air oxidation on both the inner and outer surfaces which were either pickled or blasted. In the thermogravimetric test, the hydriding rates of bare cladding specimens (no oxide prefilm) were in the range 0.9-1.6 mg/cm² h with little effect of the surface treatment. Incubation times were less than 1 h or even zero. In the tube-burst test, immediate and extensive hydrogen uptake was observed for these non-coated specimens. On the other hand, the cladding specimens with oxide prefilm exhibited lower hydriding rates ranging from 0.01 to 0.05 mg/cm² h and incubation times increased to 42 h. In addition, no hydrogen uptake was observed for all oxide-coated specimens for 100-750 h.     

Structure of zirconium alloy oxides formed in pure water studied with synchrotron radiation and optical microscopy: relation to corrosion rate

A detailed study was undertaken of oxides formed in 360 °C water on four Zr-based alloys (Zircaloy-4, ZIRLO™,¹ Zr-2.5%Nb and Zr-2.5%Nb-0.5%Cu) in an effort to relate oxide structure to corrosion performance. Micro-beam X-ray diffraction was used along with transmitted light optical microscopy to obtain information about the structure of these oxides as a function of distance from the oxide-metal interface. Optical microscopy revealed a layered oxide structure in which the average layer thickness was inversely proportional to the post-transition corrosion rate. The detailed diffraction studies showed an oxide that contained both tetragonal and monoclinic ZrO₂, with a higher fraction of tetragonal oxide near the oxide-metal interface, in a region roughly corresponding to one oxide layer. Evidence was seen also of a cyclic variation of the tetragonal and monoclinic oxide across the oxide thickness with a period of the layer thickness. The results also indicate that the final grain size of the tetragonal phase is smaller than that of the monoclinic phase and the monoclinic grain size is smaller in Zircaloy-4 and ZIRLO than in the other two alloys. These results are discussed in terms of a model of oxide growth based on the periodic breakdown and reconstitution of a protective layer.