Hydride Blisters Research Articles

Microstructure and mechanical properties of hydride blisters formed on Zircaloy-4 claddings

To assess the microstructural and mechanical characteristics of hydride blisters formed on Zircaloy-4 tubes, we employ the electron backscatter diffraction (EBSD) analyses and micro-cantilever failure tests directly on the hydride blister (HB) area. In the HB area, more than 70% of the blister area is characterized to be δ-hydride, and the HB area does not favor the orientation relation ((101¯1)α//(001)δ, [12¯10]α//[1¯10]δ) that is otherwise well-matched in the Zr matrix area. The geometrically necessary dislocation density is observed in the entire blister area, indicating a large amount of strain throughout the blister area. Also, the deformation strain is concentrated in both the hydride and α-Zr phases of HB areas; however, the hydride underwent intensive deformation in the Zr matrix. The average fracture toughness of the hydride blister area obtained through the micro-cantilever test is 1.81 MPam, indicating a brittle fracture that may threaten the integrity of spent nuclear fuel claddings.

Phase-Field Model of Hydride Blister Growth Kinetics on Zirconium Surface

The precipitation of zirconium hydride blisters on the surface of zirconium alloy cladding destroys the integrity of the material and leads to material failure, which has serious potential safety hazards. To study the corrosion kinetics of zirconium hydride blisters, we established a phase-field model coupled with anisotropic elastic, which can reflect the microstructure evolution process of zirconium hydride blisters. The model studies the natural growth of hydride blister and the structural evolution process after applying radial stress and discusses the corrosion characteristics, stress distribution, and displacement changes. Zirconium hydride blisters tend to be semi-ellipsoidal in natural growth. Applying radial outward stress will promote the growth of blisters and aggravate the harm of corrosion. The stress state changes with applying stress, which affects the elastic driving force, resulting in the change of blister morphology. This work is helpful to understand the corrosion kinetic mechanism of hydride blisters.

High temperature hardness testing of [formula omitted]-zirconium hydride: Yield stress estimation by analytical and numerical models

The mechanical properties of δ-zirconium hydrides were studied using a combination of nanoindentation techniques and numerical and analytical models. Two different alloys, containing different forms of δ-zirconium hydride (blister and rim), were analyzed at room temperature and at elevated temperatures up to 300 ∘C. A reduction of hardness from RT to 300 ∘C showed an approximately linear relationship with the change in temperature. For the δ-zirconium hydride rim, the hardness reduced from 3.73GPa at RT to 1.49GPa at 300 ∘C, a reduction of 61%. For the δ-zirconium hydride blister, hardness decreased from 3.5GPa at RT to 1.91GPa at 300 ∘C, a reduction of 45%. The yield stresses were calculated using two approaches: a numerical model proposed by Johnson to calculate values at RT, and another proposed by Dao for values up to 300 ∘C. At RT, the yield stress for the δ-zirconium hydride rim obtained was 883±21MPa, 224MPa at 300 ∘C. The δ-zirconium hydride blister showed a reduction from 879.5±10.5MPa at RT to 324MPa at 300 ∘C, a reduction in 61% and 45% for the hydride rim and blister, respectively. Through this combination of experimental and modeling techniques, this study showed the variation of hardness and yield stress in complex systems, and in different environments, using nanoindentation as the main characterization technique.

A surrogate model for the 3D prediction of in-service deformation in CANDU® fuel channels

In CANDU reactors, a contact between the Pressure Tube (PT) and the Calandria Tube (CT) could cause the development of hydride blisters that may lead to delayed hydride cracking of PTs. To maintain PT integrity, contact between these two tubes must be avoided. Under normal operating conditions (NOCs), the PT-CT contact can occur as a result of in-reactor deformation due to irradiation induced creep, irradiation growth and thermal creep of the fuel channel (FC) assembly. Numerical models have been developed to predict the time of PT-CT contact in a channel, which is required to ensure the operability of the channel and the plan for maintenance. Since the prediction of time to contact is influenced by various uncertainties, such as change in, (i) the dimensions of the FC, and (ii) the material properties and boundary conditions of the FC, probabilistic simulation-based methods have been developed to assess the PT-CT contact risk and establish adherence with provisions of the Canadian Standards Association (CSA) Standard N285.8. Numerical simulations based on a 3-dimensional finite element model (3D FEM) are essential in predicting the dimensional changes occurring during in-reactor deformation of single FCs because, unlike 1-dimensional FEM, 3D analyses correctly take into account the triaxial loading conditions in the calculation of the complete strain tensor. However, 3D FEM analyses are not practical for probabilistic assessments of an entire reactor core with 380 or 480 FCs. This paper (i) outlines the steps that are required to develop a 3D model of a FC using the ABAQUS commercial code along with user subroutines that incorporate the constitutive equations describing the in-reactor deformation of FC materials, and (ii) proposes a new surrogate model of a much simpler analytical form, which can replace the 3D FEM for assessing PT-CT contact in carrying out whole core probabilistic assessments. The proposed surrogate model not only simplifies the FE analysis and provides a more rational basis for probabilistic assessments, but most importantly allows the inclusion of diametral expansion and wall thinning in the prediction of PT-CT gap. This simple function representation gives a cost-effective and computationally efficient approach to probabilistic analysis of FCs.

Evaluation of Concentric Tube Model for Pressure Tube to Calandria Tube Gap Measurement

The CANDU® nuclear reactor fuel channels consist of a pressure tube (PT) contained within a calandria tube (CT). The gap between the hot (~300° C) PT and cooler (~50° C) CT is required since the contact could lead to hydride blister formation with subsequent cracking. This gap is measured using a drive-receive eddy current technology, providing assurance that the contact is not imminent. The analytical models of probe response to gap use flat plates to approximate the PT and CT geometry. In this paper, a semi-analytical model that approximates the PT within the CT geometry as two concentric tubes with correct PT curvature, but a CT diameter, which is adjusted to change gap, is compared with a flat-plate model of probe response to changing gap against experimental measurements using standard CT and PT samples, with varying wall thickness and resistivity that simulates in-reactor variation of these parameters. The concentric and flat plate models both predict gap values with an average error of 0.1 mm between contact and 5 mm gap at 4.2 kHz and 8.0 kHz. The concentric model has an average relative gap error at the maximum gap of 2% at 4.2 kHz and 8.0 kHz, whereas that of the flat-plate model is 10% and 12%, respectively. The improved accuracy of the concentric model at larger gaps is attributed to the incorporation of tube curvature, which limits the probe’s magnetic field spread when compared to the flat-plate model.

Delayed hydride cracking of Zr-2.5%Nb pressure tube material due to partially constrained precipitates

Formation of partially constrained precipitates such as hydride blisters and oxide nodules have been reported on surfaces of Zr-alloy components of pressurised heavy water reactors and is associated with a large increase in volume. Such a change in volume imposes large stresses in the material surrounding the precipitate and may facilitate stable crack growth through delayed hydride cracking. In this work, the stress field of the partially constrained precipitates with different depth and aspect ratio has been computed using a finite element method. The computed stress field is used to predict the region in the matrix in which radial hydride is likely to form and fracture, by taking into consideration grain-size, texture and multi-axial state of stress. For a hypothetical crack just below the precipitate, stress intensity factors are estimated using material properties for both unirradiated and irradiated pressure tube materials. The results are compared with the threshold stress intensity factors required for crack growth due to delayed hydride cracking.

Self‐inflicted fracture of expanding surface precipitates

AbstractThis work concerns spontaneous fracture of growing brittle precipitates in an elastic plastic matrix. The mass of the precipitate is increasing as more transformed matrix material is added to it. Under stress‐free conditions, the new phase occupies a larger volume than the original matrix material. Just outside the expanding precipitate, the matrix undergoes stretching beyond the elastic limit. The influence of the elastic plastic material behaviour is studied. A phase field model that keeps track of the phase composition is used. Both cases with a crack and without a crack are included. The growth histories from microscopic to macroscopic precipitate sizes are followed. Growth of the precipitate is very slow and quasi‐static mechanical equilibrium is assumed at all time. The result is compared with observations of hydride blisters that are formed on surfaces of zirconium alloys. The numerical model is qualified against a derived exact solution for a cylindrical precipitate without a crack. The numerical result predicts a position of the growing crack that is confirmed by the observations. Also, the predicted length of the crack is in fair agreement with the experimental observations. The depth of the blister is slightly larger than what is found at the experiments. Also, it is found that the incorporated transformed phase rejects the compression, which creates an increasing tensile stress in the inner part of the precipitate.

Understanding the mechanics of creep deformation to develop a surrogate model for contact assessment in CANDU® fuel channels

A key element of the fuel channel life cycle management in CANDU® reactors is to prevent contact between the pressure tube (PT) and the calandria tube (CT) and to avoid the development of hydride blisters that lead to delayed hydride cracking of the PT. The PT-CT contact is the result of in-reactor deformation due to irradiation induced creep of the fuel channel assembly, which in turn is affected by uncertainties associated with various parameters like material properties, dimensional changes in the channel and boundary conditions (e.g., end slopes) of the channel. To account for these uncertainties, probabilistic assessment methods are developed to evaluate the risk of PT-CT contact and demonstrate compliance with provisions of the CSA Standard N285.8. Currently, a simulation is based on probabilistic assessments in which input parameters to a finite element model (FEM) of creep deformation are randomly sampled from their respective distributions. A simulation model involves numerous repetitive solutions of the FEM model to determine the probability distribution of PT-CT gap and the time to contact. Since the creep deformation analysis using FEM is computationally involved, this brute force Monte Carlo simulation method is not an efficient way to carry out the probabilistic assessment of the reactor core. This paper proposes a new line of thinking for probabilistic assessments of PT-CT contact in fuel channels, which is based on replacing a full FEM model by a surrogate model of a much simpler analytical form. The surrogate model not only simplifies the creep deformation analysis, but also provides a more logical basis for probabilistic assessments. This paper presents an insightful analysis of creep deformation and shows that a simple surrogate model can be developed to predict the PT-CT gap as a linear function of two primary random variables, namely, a creep factor and end slopes. This simplified representation has a far reaching effect on the probabilistic assessment of fuel channels.

Effects of Zr-hydride distribution of irradiated Zircaloy-2 cladding in RIA-simulating pellet-clad mechanical interaction testing

Abstract A series of simulated reactivity-initiated accident (RIA) tests on irradiated fully recrystallized boiling water reactor Zircaloy-2 cladding has been performed by means of the expansion-due-to-compression (EDC) test method. The EDC method reproduces fuel pellet–clad mechanical interaction (PCMI) conditions for the cladding during RIA transients with respect to temperature and loading rates by out-of-pile mechanical testing. The tested materials had a large variation in burnup and hydrogen content (up to 907 wppm). The results of the EDC tests showed variation in the PCMI resistance of claddings with similar burnup and hydrogen content, making it difficult to clearly identify ductile-to-brittle transition temperatures. The EDC-tested samples of the present and previous work were investigated by light optical and scanning electron microscopy to study the influence of factors such as azimuthal variation of the Zr-hydrides and the presence of hydride rims and radially oriented hydrides. Two main characteristics were identified in samples with low ductility with respect to hydrogen content and test temperature: hydride rims and radial hydrides at the cladding outer surface. Crack propagation and failure modes were also studied, showing two general modes of crack propagation depending on distribution and amount of radially oriented hydrides. It was concluded that the PCMI resistance of irradiated cladding under normal conditions with homogenously distributed circumferential hydrides is high, with good margin to the RIA failure limits. To further improve safety, focus should be on conditions causing nonfavorable hydride distribution, such as hydride reorientation and formation of hydride blisters at the cladding outer surface.

Phase analysis of hydride blister in zirconium alloy

This paper mainly concerns phase analysis of hydride blister and phase transformation at elevated temperature with Zircaloy-4 cladding. With electron diffraction patterns from the blister sample, single δ phase hydride is selected, especially, to investigate the phase transformation of δ-hydride with TEM. Under the rapid cooling condition, single δ-hydride transforms to heterogeneous ultrafine lamellar-like structure with the δ and γ-hydride phases. However, the phase transformation from δ-hydride to γ-hydride has not been reported around δ-phase domain in the Zr-H phase diagram, and thus, we suggest that the γ-hydride is able to be generated in the δ phase domain as well as in the α+δ phase domain. It is also proven that the γ phase tends to transform to the stable δ phase with sufficient holding time at the elevated temperature during in-situ TEM experiment.

Microstructure characterization of a hydride blister in Zircaloy-4 by EBSD and TEM

This paper presents a detailed microstructural characterization by EBSD and TEM techniques of a hydride blister grown on a Zircaloy-4 plate. The hydrides formed are found to be of δ-type, and exhibit an increasing volume fraction from alloy matrix, across matrix-blister boundary, and into the blister. The microstructure of the blister mainly consists of δ-hydride platelets, however there is still un-transformed α-Zr phase inside the blister. The orientation relationship between the hydride and alloy matrix has been analyzed. In addition to the commonly observed (0001)α−Zr//{111}δ orientation relationship, we found hydrides with a secondary (0001)α−Zr//(100)δ orientation relationship. A correlation was drawn between the formation of hydrides with the two orientation relationships and the parent grain orientation. TEM observation revealed the distribution of the remnant α-Zr phase inside the blister, which had either a nodular shape or appeared as thin nano-sized layers among the hydride platelets. The microstructure of the hydrides has been characterized by HRTEM, exhibiting a uniform atomic structure with high dislocation density. Lastly, δ-hydride embryos obeying the primary orientation relationship were found within untransformed α-grains, indicating a consistent nucleation mechanism during blister growth.

Orientation relationships between α-zirconium and δ-hydride within a hydride blister

Both the expected and an additional orientation relationship between α-Zr and δ-hydride in blistered zirconium alloys are explored through the reconstruction of the parent α-Zr phase from electron backscatter diffraction maps of δ-hydride. Parent and child variant relationships for the transformation are presented with the aim of reconstruction of the parent α-Zr grain structure and texture from orientation maps of the δ-hydride at varying distances from the blister centre in a recrystallized Zircaloy-4 sample. Up to 13% of the δ-hydride is found to be variants of the additional orientation relationship, the fraction of which decreases with increasing distance from the blister centre. Texture reconstructions by other experimental methods are accordingly suggested to incorporate the additional orientation relationship.

Obtention of the constitutive equation of hydride blisters in fuel cladding from nanoindentation tests

It is well known that the presence of hydrides in nuclear fuel cladding may reduce its mechanical and fracture properties. This situation may be worsened as a consequence of the formation of hydride blisters. These blisters are zones with an extremely high hydrogen concentration and they are usually associated to the oxide spalling which may occur at the outer surface of the cladding. In this work, a method which allows us to reproduce, in a reliable way, hydride blisters in the laboratory has been devised. Depth-sensing indentation tests with a spherical indenter were conducted on a hydride blister produced in the laboratory with the aim of measuring its mechanical behaviour. The plastic stress-strain curve of the hydride blister was calculated for first time by combining depth-sensing indentation tests results with an iterative algorithm using finite element simulations. The algorithm employed reduces, in each iteration, the differences between the numerical and the experimental results by modifying the stress-strain curve. In this way, an almost perfect adjustment of the experimental data was achieved after several iterations. The calculation of the constitutive equation of the blister from nanoindentation tests, may involve a lack of uniqueness. To evaluate it, a method based on the optimization of parameters of analytical equations has been proposed in this paper. An estimation of the error which involves this method is also provided.

Phase Field Modelling of Formation and Fracture of Expanding Precipitates

It is a common belief that embedded expanding inclusions are subjected to an internal homogeneous compressive hydrostatic stress. Still, cracks that appear in precipitates that occupy a larger volume than the original material, are frequently observed. The appearance of cracks has since long been regarded as a paradox. In the present study it is shown that matrix materials that increases its volume even several percent during the precipitation process develop a tensile hydrostatic stress in the centre of the precipitate. This is the result of a complicated mechanical-chemical phase transformation process. The process is here studied using a Landau phase feld model. Before the material is transformed and incorporated in a precipitate it undergoes stretching beyond the elastic strain limit because of the presence of already expanded material. During the phase transformation, the accompanying volumetric expansion cannot be fully accommodated which instead creates an internal compressive stress and adds tension in the surrounding material. As the growth of the precipitate proceeds, a region with increasing tensile stress develops in the interior of the precipitate. This is suggested to be the most probable cause of the observed cracks. First the mechanics that lead to the tension is computed. The infuence of elastic-plastic properties is studied both for cases both with and without cracks. The growth history from microscopic to macroscopic precipitates is followed and the result is compared with observations of so called hydride blisters that are formed on surfaces of zirconium alloys in the presence of hydrogen. A common practical situation is when the zirconium is in contact with an object of lower temperature. Then the cooled spot attracts hydrogen that make the zirconium transform to a metal hydride with the shape of a blister. The simulations predicts a final size and position of the growing crack that compares well with the experimental observations.

Zirconium hydrides and Fe redistribution in Zr-2.5%Nb alloy under ion irradiation

Zr-2.5%Nb alloy is used to fabricate the pressure tubes of the CANDU reactor. The pressure tube is the primary pressure boundary for coolant in the CANDU design and is susceptible to delayed hydride cracking, reduction in fracture toughness upon hydride precipitation and potentially hydride blister formation. The morphology and nature of hydrides in Zr-2.5%Nb with 100 wppm hydrogen has been investigated using transmission electron microscopy. The effect of hydrides on heavy ion irradiation induced decomposition of the β phase has been reported. STEM-EDX mapping was employed to investigate the distribution of alloying elements. The results show that hydrides are present in the form of stacks of different sizes, with length scales from nano- to micro-meters.Heavy ion irradiation experiments at 250 °C on as-received and hydrided Zr-2.5%Nb alloy, show interesting effects of hydrogen on the irradiation induced redistribution of Fe. It was found that Fe is widely redistributed from the β phase into the α phase in the as-received material, however, the loss of Fe from the β phase and subsequent precipitation is retarded in the hydrided material. This preliminary work will further the current understanding of microstructural evolution of Zr based alloys in the presence of hydrogen.

Quantification of dislocations densities in zirconium hydride by X-ray line profile analysis

Zirconium-based components in nuclear power plants are embrittled by precipitates of δ zirconium hydride, which involves a martensitic-type transformation of the hexagonal α-Zr lattice into the face-centered cubic Zr sublattice of the hydride. As a result, the hydride precipitates have a complex and heavily distorted internal structure that manifests as broad peaks in X-ray diffraction experiments. By a detailed analysis of the peak widths measured for different crystal planes we have found that most of this broadening is the result of dislocations. The analysis also showed that δ-hydride has very anisotropic mechanical elastic properties, in agreement with ab-initio simulations presented in the literature.Provided with this peak-broadening model, we have quantified dislocation densities within δ-hydrides precipitated in several Zr alloys, by analyzing previously published X-ray diffraction experiments performed at three synchrotron X-ray sources. The specimens investigated correspond to components affected by different hydride embrittling processes, namely: (i) samples from various components, charged in the laboratory with H contents in the ∼250 wt ppm range, (ii) laboratory-produced hydride blisters in Zr2.5%Nb pressure tubes; and (iii) Zircaloy-4 specimens machined from cooling channels of Atucha I nuclear power plant after 10 years in-service, containing ∼140 wt ppm of equivalent H content and subjected to an estimated fast neutron fluence of ∼1022 neutrons/cm2. Results show that dislocations densities in the δ-hydrides are large (5–20 × 1015 cm−2) and vary among the different specimens. We also found that dislocations densities in the hydride are proportional to the fraction of hydrides already formed in the matrix, which was interpreted as the effect of matrix hardness in the precipitate structure.

2D simulation of hydride blister cracking during a RIA transient with the fuel code ALCYONE

This paper presents 2D generalized plain strain simulations of the thermo-mechanical response of a pellet fragment and overlying cladding during a RIA transient. A fictitious hydride blister of increasing depth (25 to 90% of the clad thickness) is introduced at the beginning of the calculation. When a pre-determined hoop stress is exceeded at the clad outer surface, radial cracking of the blister is taken into account in the simulation by a modification of the mechanical boundary conditions. The hoop stress criterion is based on Finite Element simulations of laboratory hoop tensile tests performed on highly irradiated samples with a through-wall hydride blister. The response of the remaining clad ligament (beneath the cracked blister) to the pellet thermal expansion is then studied. The simulations show that plastic strains localize in a band orientated at ∼45° to the radial direction, starting from the blister crack tip and ending at the clad inner wall. This result is in good agreement with the ductile shear failures of the clad ligaments observed post-RIA transients. Based on a local plastic strain failure criterion in the shear band, ALCYONE simulations are then used to define the enthalpy at failure in function of the blister depth.

A quantitative phase field model for hydride precipitation in zirconium alloys: Part II. Modeling of temperature dependent hydride precipitation

A quantitative free energy functional developed in Part I (Shi and Xiao, 2014 [1]) was applied to model temperature dependent δ-hydride precipitation in zirconium in real time and real length scale. At first, the effect of external tensile load on reorientation of δ-hydrides was calibrated against experimental observations, which provides a modification factor for the strain energy in free energy formulation. Then, two types of temperature-related problems were investigated. In the first type, the effect of temperature transient was studied by cooling the Zr–H system at different cooling rates from high temperature while an external tensile stress was maintained. At the end of temperature transients, the average hydride size as a function of cooling rate was compared to experimental data. In the second type, the effect of temperature gradients was studied in a one or two dimensional temperature field. Different boundary conditions were applied. The results show that the hydride precipitation concentrated in low temperature regions and that it eventually led to the formation of hydride blisters in zirconium. A brief discussion on how to implement the hysteresis of hydrogen solid solubility on hydride precipitation and dissolution in the developed phase field scheme is also presented.

Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding

It is well known that the mechanical properties of the nuclear fuel cladding may be affected by the presence of hydrides. The average mechanical properties of hydrided cladding have been extensively investigated from a macroscopic point of view. In addition, the mechanical and fracture properties of bulk hydride samples fabricated from zirconium plates have also been reported. In this paper, Young’s modulus, hardness and yield stress are measured for each phase, namely zirconium hydrides and matrix, of pre-hydrided nuclear fuel cladding. To this end, nanoindentation tests were performed on ZIRLO samples in as-received state, on a hydride blister and in samples with 150 and 1200ppm of hydrogen homogeneously distributed along the hoop direction of the cladding. The results show that the measured mechanical properties of the zirconium hydrides and ZIRLO matrix (Young’s modulus, hardness and yield stress) are rather similar. From the experimental data, the hydride volume fraction in the cladding samples with 150 and 1200ppm was estimated and the average mechanical properties were calculated by means of the rule of mixtures. These values were compared with those obtained from ring compression tests. Good agreement between the results obtained by both methods was found.

Formation and characterization of hydride blisters in Zircaloy-4 cladding tubes

This article is focused on the formation of hydride blisters in zirconium alloys an experimental and theoretical standpoint, and their characterization in terms of morphology, hydrides crystallographic phases, hardness and hydrogen concentration. An experimental setup was developed to grow hydride blisters on pre-hydrided Zircaloy-4 cladding tubes by thermo-diffusion. The thermal conditions were optimized based on thermo-diffusion calculations, that take into account the hysteresis in the hydrogen solubility limit, to obtain a high blister growth rate. Micro-X-ray Diffraction (XRD), nano-hardness and Elastic Recoil Detection Analysis (ERDA) showed that the blisters contain a hydrogen gradient, with pure δ-hydride phase close to the external surface over one third of the blister depth. Thermo-diffusion calculations showed these half thickness blisters should grow in only a few days in PWR conditions. Eventually, the Diffusion Equilibrium Threshold (DET) was defined as a criterion that limits the blister growth, and emphasizes that the hysteresis in the hydrogen solubility limit in zirconium must be taken into account to model hydrogen thermo-diffusion in zirconium alloys.