Alloy Pressure Tubes Research Articles

Structural integrity assessment of CANDU pressure tubes using Sobol indices for global sensitivity analysis

CANDU11CANDU is a registered trademark of Atomic Energy of Canada Limited. reactors are a channel-type design where up to 480 zirconium alloy pressure tubes (PTs) act as the reactor pressure boundary. Fitness-for-service standards such as CSA N285.8 (CSA Group, 2015) have been established to prescribe the requirements of pressure tube evaluations to maintain their integrity throughout their lifetime. Probabilistic leak-before-break (LBB) assessments can be used in evaluations to demonstrate that the likelihood of a flaw growing past the stability point while undetected (or before a safe shutdown can be achieved) is below regulatory limits. The assessments typically require a ranking of model parameters based on their influence. The ranking process traditionally uses expert judgment and local sensitivity analysis methods. There has been recent interest in implementing global methods for LBB assessments as they can manage large numbers of variables and extensive sampling spaces. However, their adoption is hindered by the computational cost of the models involved.The current study proposes a method for performing global sensitivity parameter rankings during probabilistic LBB assessments. The study is the continuation of the CANDU nuclear power plant (NPP) pressure tube (PT) case study in El Bouzidi et al., (2024) where a probabilistic approach was implemented in RAVEN (Rabiti and Alfonsi, 2019) to estimate whether the probability of unstable crack growth in PTs is within acceptable limits. This follow-up work proposes surrogate modeling using a high-dimensional model representation to derive Sobol indices and establish a parameter influence ranking. The methodology is demonstrated in a Monte Carlo simulation campaign of a pressure tube inlet rolled joint. The implemented approach successfully identified and quantified critical parameters of the model while efficiently ranking and quantifying the interactions between input variables.

The radial delayed hydride cracking behavior of Zr-2.5Nb alloy pressure tube at different temperatures

Cantilever beam samples with V-notches were employed to measure the radial delayed hydride cracking rate (DHCR) in Zr-2.5Nb alloy pressure tubes within the temperature range from 120 °C to 250 °C in this study. Both initiation and propagation of DHC were monitored using the direct current potential drop (DCPD) technique. The results show that radial DHCR and temperature fit the Arrhenius relationship within the temperature as mentioned above range, yielding a crack propagation activation energy of 38.66 kJ/mol. It was observed that the incubation periods for radial DHC initiation typically shortened with rising temperature from 180 °C to 250 °C, and the anisotropy of DHCR is also discussed in this study. Besides, the presence of undissolved circumferential hydrides appears to affect radial DHC behavior.

Microstructure and fracture toughness of 850°C quenched and aged Zr-2.5Nb alloy

The Zr-2.5Nb alloy pressure tubes manufactured by Water Quenched and Aged (WQA) fabrication route has reportedly exhibited superior resistance against in-reactor dimensional changes. WQA route comprises three consecutive thermo-mechanical treatments viz., solution heat treatment (SHT), cold working and vacuum aging, which govern the final microstructure and therefore fracture toughness of the pressure tube material. This work describes the influence of SHT soaking duration, degree of cold working and aging temperature on the microstructure and fracture toughness of WQA Zr-2.5Nb alloy. WQA pressure tubes are expected to have better resistance to in-reactor dimensional changes. However, if proper care of hydrogen absorption during thermo-mechanical processing is not taken, though resistance to creep will improve expectedly but fracture toughness at ambient temperature will deteriorate.

Effect of temperature on the delayed hydride cracking rate of Zr-2.5Nb alloy pressure tubes

Delayed hydride cracking (DHC) is a potential failure mechanism of Zr-2.5Nb alloy pressure tubes, employed in heavy-water reactors. This study focused on understanding the DHC behavior of the Zr-2.5Nb pressure tubes within the temperature range 393–523 K. Our findings demonstrate that within this temperature range, the delayed hydride cracking rate (DHCR) and temperature follow the Arrhenius relationship, with the activation energy of crack propagation is found to be 43.16 kJ/mol. The changes in the fracture morphologies at different temperatures are analyzed. Variations in the fracture hydride lengths with temperature and the stress intensity factor K are obtained. Furthermore, it is found that the influencing mechanism of temperature on DHC behavior changes as the temperature increases.

Numerical Study of the Hydride Embrittlement in Zirconium Alloy using XFEM

A numerical model for hydride embrittlement in Zirconium alloy (Zr–2.5Nb) is developed utilizing the extended finite element method (XFEM). Hydride embrittlement reduces the ductility and failure time of a metal/alloy. During hydride embrittlement, stress-directed hydrogen diffusion, metal-hydride phase transformation, mechanical deformation, and hydride precipitation occur simultaneously. The present model incorporates all these processes and is able to predict the hydrogen concentration and the hydride fraction distribution under any externally applied stress field. In this work, both the steady and transient hydrogen diffusion cases are evaluated. Further, the XFEM is utilized to develop a model of hydride embrittlement in the presence of a crack. The first step of the hydride embrittlement process is the diffusion of hydrogen. According to Fick’s law of diffusion, hydrogen diffusion is directly dependent on hydrostatic stresses and hydrogen concentration gradient under external stresses. The next step is the hydride precipitation in hydride embrittlement, where the expansion of material takes place that changes the hydrostatic stress field. Thus, studying the effect of precipitation of hydride on hydrostatic stresses is essential. Moreover, the process of hydride embrittlement is highly influenced by residual stresses in the structure. Hence, the effect of residual stress present in the zirconium alloy pressure tube (PT) is also evaluated. The results indicate that the residual tensile stresses contribute to the growth of hydride, which will reduce the material failure time.

Anisotropy study of hydrogen diffusion along different directions of Zr-2.5%Nb alloy pressure tube using neutron imaging

Hydrogen diffusion studies in Zr-2.5% Nb alloy pressure tube material are important inputs for modeling localized embrittlement phenomenon manifested in pressure tubes. The microstructure of the pressure tube consists of lamellar α-Zr along with the β-phase present as stingers between two alpha laths as well as fine and coarse beta globules. This two-phase microstructure along with a vast difference in diffusivities for the alpha and beta phase necessitates the need for evaluating the diffusion parameters along the three orthogonal directions of the pressure tube. This work focuses on studying the diffusivity values for hydrogen in Zr-2.5%Nb alloy along different pressure tube directions in the temperature range is 250 ˚C -450 ˚C and also correlates it to the microstructure. FEM studies have also been carried out to identify the concentration region that is unaffected by stress imposed due to the presence of a hydride layer. Neutron radiography has been used to study diffusion along the axial, circumferential and radial directions of an as-received un-irradiated pressure tube. It was observed that the diffusion along the axial and circumferential directions is similar while it is slightly lower in the case of radial direction. A theoretical approach has also been followed for evaluating the diffusivity values for the α-Zr and β-Zr phases for the same temperature range.

Anisotropy and variability in thermal creep behaviour of Zr-2.5Nb pressure tube

The Zr-2.5Nb alloy pressure tubes of Indian Pressurized Heavy Water Reactors were manufactured from double melted ingots initially and subsequently from quadruple melted ingots to reduce impurities such as chlorine, phosphorus and carbon. In the current investigation, creep tests were carried out in the stress range of 0.7–0.9 times the yield strength and the temperature range of 350 °C–450 °C to characterize the thermal creep behaviour of Zr-2.5Nb alloy pressure tubes fabricated from double and quadruple melted ingots. The rupture time, minimum creep rate, stress exponent, activation energy and threshold stresses were obtained by carrying out the creep tests of the samples fabricated with their axis parallel to the axial and transverse directions of the pressure tubes. Experimental analysis revealed that the rupture time for double melted tube was 2–3 times lower than that of quadruple melted tubes, whereas no significant change in minimum creep rate was observed. Stress exponent values were obtained in the range of 3–5 signifying dislocation creep as the dominant creep mechanism. To predict the long-term creep properties, the master curves employing the Larson-Miller and Monkman-Grant methods were also evaluated. The effect of microstructural anisotropy, alloying elements and impurities on the thermal creep behaviour of Zr-2.5Nb pressure tube has been investigated.

Determination of relative sensitivity factor, sputtering rate, and detection limits of deuterium in deuterium ion‐implanted Zircaloy‐4 using secondary ion mass spectrometer

Zirconium‐based alloys are extensively used in the nuclear industry as materials for reactor core components. These alloys are used in fuel cladding and pressure tubes with 1H2O or 2H2O as the primary heat exchange fluid. Continuous usage of pressure tubes and fuel cladding inside the reactor results in heavy ingress of hydrogen (1H) or deuterium (2H), which gives rise to the formation of hydrogen (or deuterium) precipitates in the host of zirconium matrix. This is called hydrogen (or deuterium) embrittlement phenomenon. Therefore, determination of hydrogen or deuterium concentration periodically in zirconium alloy pressure tubes and cladding are crucial for safe operation of nuclear reactors. Hence, in this work, secondary ion mass spectrometry (SIMS) has been utilized to determine the relative sensitive factor (RSF), sputtering rate, and detection limits of 2H in 2H+ ion‐implanted Zircaloy‐4 sample. SIMS analysis has been carried out using O2+ and Cs+ primary ion beams. Depth profiles of various elemental and molecular secondary ions have been monitored to determine the RSF values and detection limits of 2H in the sample under different primary ion beam analysis conditions. Sputtering rate under different analysis conditions has also been determined from the correlation between the projected range of 2H+ ions in Zircaloy‐4 sample determined from Stopping and Range of Ions in Matter (SRIM) software and the depth of the crater formed by primary ion beam bombardment process. The projected range of 2H+ ions in Zircaloy‐4 was estimated to be ~524.2 nm using the simulation carried out by SRIM software. The projected range, evaluated by SRIM software, was utilized for depth scale calibration of SIMS sputtered time profiles. The SIMS depth profiles were also calibrated by determining the depth of the SIMS crater by using optical depth profilometer. The values of range estimated by SRIM as well as optical profilometer were found to be in good agreement, within instrument uncertainty.

Effect of manufacturing route on thermal creep behaviour of Zr-2.5Nb pressure tube alloy used in Indian PHWR

Indian Pressurized Heavy Water Reactors (IPHWRs) uses Zr-2.5Nb alloy pressure tube (PT) in cold work and stress relieved condition. These PTs are subjected to extreme conditions of stress, temperature and irradiation. Since past three decades of PHWR operation in India, fabrication routes of PTs for IPHWR has undergone several modifications to enhance its in-reactor performance. The deformation of PT in PHWR occurs due to irradiation creep, irradiation growth and thermal creep. In the present study, the different generations of PTs used in IPHWR are tested for its thermal creep behaviour. The accelerated creep rupture tests were carried out using flat tensile specimens with its axis parallel to the longitudinal direction of the PT in the temperature range of 350 to 450°C under stresses in the range of 0.7 to 0.9 times of yield strength at the respective temperatures. The minimum creep rates were evaluated along with the stress exponents and activation energies. The values of stress exponent and activation energy revealed the underlying creep mechanisms responsible for the creep deformation. Microstructural examination was carried out to understand the need and efficacy of the new fabrication route. The observed thermal creep behaviour of all the generations of pressure tubes were rationalised in terms of their respective texture and microstructures.

Analysis of Dimensional Change in Zr-2.5%Nb CANDU Pressure Tube Material aged at 300-400°C using Neutron Diffraction

The primary boundary of the CANDU reactor consists of Zr-2.5%Nb alloy pressure tubes and SA106 alloy feeder pipes. The pressure tubes are exposed to fast neutron (E>1Mev) irradiation during reactor operation, and experience dimensional changes. The actual dimensional change in the pressure tube during in-service operation is greater than the design estimate. It is not fully understood what is responsible for the dimensional changes in the pressure tubes. In this study, to accelerate the aging effect on the pressure tube material, aging was performed at 300-400°C for up to 20,000 hours and neutron diffraction was used to track variation in lattice spacing, thereby identifying the reason for the dimensional change. The analysis result showed that α-Zr, the main phase of Zr-2.5%Nb, had little dimensional change at 350°C or lower. On the other hand, the aging treatment at 400°C resulted in anisotropic expansion in the α-Zr of the HCP crystal, of 0.02% and 0.08% in the (1010) and (0002) directions, respectively. It was confirmed that above 350oC the β-phase in the Zr-2.5%Nb alloy was decomposed to precipitate β-Nb . The driving force of the lattice expansion may be due to the diffusing out tendency of the super saturated Nb in α-Zr and due to the residual cold working effect applied during the manufacturing process. The enhancing effects of fast neutron irradiation on lattice diffusion are discussed in detail.

Isothermal stress reorientation of hydrides in Zr-2.5Nb pressure tube alloy

• For the first time, in-situ hydride reorientation phenomenon in Zr-2.5Nb alloy has been investigated. • Threshold stress for reorientation of hydrides was determined under isothermal condition using a specially designed setup. • Threshold stress values determined using in-situ and ex-situ methods were comparable. The precipitation of hydrides normal to the transverse direction in Zr-2.5%Nb pressure tube alloy when cooled under stress above a threshold value, is known as stress reorientation of hydrides. The ex-situ threshold stress values reported in the literature were determined using pre-charged samples heated to a peak temperature to dissolve all the hydrides, cooled to reorientation temperature followed by cooling under stress from reorientation temperature. Leger and Donner performed in-situ reorientation experiments using samples having hydride deposited on its surface. In both cases, hydride reorientation occurred during cooling. In the present investigation, for the first time, experiments were done to study the reorientation phenomenon using in-situ gaseous charging of samples under stress at a constant temperature using a specially designed set up. It is expected that the threshold stress determined by in-situ method used in the present investigation will more closely resemble the formation of radial hydrides under reactor operating conditions. In this work, the threshold stress values for reorientation of hydrides in Zr-2.5%Nb pressure tube material were determined using both in-situ and ex-situ methods. The precipitation of hydrides normal to the transverse direction in Zr-2.5%Nb pressure tube alloy when cooled under stress above a threshold value is known as stress reorientation of hydrides. In the present investigation, for the first time the threshold stress (σ th ) for reorientation of hydrides in Cold worked and stress relieved (CWSR) Zr-2.5Nb alloy pressure tube material was determined by using a specially designed fixture shown below consisting of load train comprising sample, pull rod and load cell connected in a series, mounted between two rods welded to a flange. A flexible bellow is welded to flange and pull rod to accommodate the displacement of pull rod and to ensure vacuum leak tightness. Ports are provided in the flange to measure the temperature of sample and to connect the system to vacuum pump. The entire setup after sample loading is inserted into the glass tube and loaded into a resistance heated furnace samples were charged with hydrogen under stress at constant temperature. It is expected that the threshold stress determined by in-situ method will more closely resemble the formation of radial hydrides under reactor operating conditions.

Effect of Specimen Thickness on Threshold Stress Intensity Factor (K IH) Associated with DHC in Zr-2.5 Nb Alloy Pressure Tube Material

Abstract Delayed hydride cracking (DHC) behavior for zirconium alloys is characterized in terms of DHC velocity, VDHC, and Threshold Stress Intensity factor, KIH. The effect of specimen thickness on KIH associated with DHC of unirradiated Zr-2.5 Nb pressure tube material was investigated. For this purpose, KI at crack arrest for 1 mm–, 2 mm–, and 4.5 mm–thick specimens was determined using the load drop technique. Significant tunneling of the crack was observed for 1 mm– and 2 mm–thick specimens in the load drop test at 250°C. The KI at crack arrest was found to be increasing with the decrease in specimen thickness. However, extended finite element model (FEM) simulation revealed that KI at crack arrest in the center of the 1-mm specimen and 2-mm specimen is comparable to the values 9.2 ± 0.5 MPa.m1/2obtained for the 4.5 mm–thick specimen. Hence, if tunneling is not present in the specimen, the KIH was found to be independent of specimen thickness.

Investigation of hydrogen diffusivity in Zr-2.5%Nb alloy pressure tube material using Metallography and Neutron Radiography

Zr-2.5%Nb alloy is used as pressure tube material in Indian pressurised heavy water reactors (PHWR). Hydrogen diffusion studies have been carried out in this alloy in the temperature range 523 K-723 K. Samples were prepared by electrolytic hydrogen charging at one end and then subsequent annealing at said temperatures to form a concentration gradient. Neutron radiography and for the first time, metallography techniques have been used for hydrogen content profiling. Calibration studies for hydrogen in Zr-2.5%Nb alloy have also been carried out at Dhruva imaging beamline to determine the absolute hydrogen content from neutron transmission values. The hydride area fraction has been used as the measure of hydrogen content in metallography studies. The absolute hydrogen content for radiography and hydride area fraction for metallography; vs distance data were fitted with error function to obtain diffusivities at various temperatures. The diffusivity data were further processed to obtain the temperature-dependent diffusion coefficients for both techniques in the temperature range of 523 K-723 K. The results obtained from both techniques are consistent, establishing the fact that a simple lab-based technique like metallography can also be used for diffusion studies.

Effect of β Phase Decomposition on Recrystallization in α-Zr Region in Zr-2.5%Nb Pressure Tube Material

The effects of β phase decomposition on recrystallization and texture variation in Zr-2.5% Nb alloy pressure tube material were investigated. Isothermal annealing was conducted at 350 to 550 <sup>o</sup>C for 240 hours, and isothermal annealing was performed at 500 <sup>o</sup>C for 240 to 3,000 hours. The recrystallization and texture variation were analyzed by inverse pole figure variation using the XRD and EBSD methods. Annealing in α-Zr region at below 610 <sup>o</sup>C induced recrystallization and texture variation in the α-Zr. These results differ from those from a previous annealing study of the α+β region at 750-830 <sup>o</sup>C. Annealing above 400 <sup>o</sup>C for 240 hours caused β-Zr decomposition into β-Nb. The decomposition of the β-phase by annealing above 475 <sup>o</sup>C caused a contraction of 7.5% in the d(110) spacing in the β-phase, and a reduction in the volume fraction of the β phase by about 80%. It seems that the stress internally formed by the lattice contraction of the β-phase provides the driving force for recrystallization. In addition, it suggests that the newly formed α-Zr produced by β phase decomposition provides new nucleation sites for recrystallization and causes texture variation in the α-Zr. The reason why the recrystallization and the texture variation occurs only in the α-Zr stable region at below 610 <sup>o</sup>C is discussed.

Phase Transformation and Atypical Variants in an Extruded Two Phase Zirconium Tube

The texture evolution of a Zr-2.5wt pctNb alloy pressure tube is characterized after the $$\alpha \rightarrow \beta $$ transformation and the $$\beta \rightarrow \alpha $$ transformation by room temperature EBSD maps and reconstruction of the high temperature $$\beta $$ phase. The $$\beta $$ phase forms a deformation texture which varies along the radial direction of the tube and changes after thermal cycling. Variant selection during the $$\beta \rightarrow \alpha $$ phase transformation is shown to be strongly influenced by residual stress within the material, with a more randomized variant selection for water quenched samples. Slower cooling rates were found to create orientations in addition to the 12 $$\alpha $$ variants expected by the Burger’s orientation relationship in a given $$\beta $$ grain.

Study on the effect of Ar9+ ion irradiation of Zr–2.5 wt.% Nb alloy pressure tube

ABSTRACTResponse of Zr–2.5 wt.% Nb alloy pressure tube, used in PHWR nuclear reactors, to 315 keV Ar9+ ion irradiation at room temperature was investigated in the fluence range of 3.1 × 1015–4.17 × 1016 Ar9+ cm−2. Changes in microstructural parameters, viz., the size of coherently scattering domains, microstrain and dislocation density, upon irradiation were ascertained through grazing incidence X-ray diffraction. In general, a decrease in domain size was observed with fluence with a corresponding increase in microstrain and dislocation density. Residual stress measurement showed the development of compressive stresses in place of tensile after irradiation. Transmission electron microscopy showed the formation of dislocation loops of ⟨a⟩-type and ⟨c⟩-type during irradiation. The hardness of irradiated samples, probed through nanoindentation technique, was found to be higher in comparison with unirradiated samples. The above findings have been rationalised on the basis of the defects generated during the Ar9+ ion irradiation.

Influence of hydride orientation on fracture toughness of CWSR Zr-2.5%Nb pressure tube material between RT and 300 °C

An experimental setup was designed, fabricated and used to form radial hydrides in Zr-2.5%Nb alloy pressure tube spool. The design of setup was based on ensuring a hoop stress in the spool greater than threshold stress for reorientation of hydrides in this alloy, which was achieved by manipulating the thermal expansion coefficient of the plunger and pressure tube material and diametral interference between them. The experimental setup was loaded on a universal testing machine (UTM) fitted with an environmental chamber and subjected to a temperature cycle for the stress reorientation treatment. The metallographic examination of the hydrogen charged spools subjected to stress re-orientation treatment using this set up revealed formation of predominantly radial hydrides. The variation of fracture toughness of material containing radial hydride with test temperature showed typical ‘S’ curve behavior with transition temperatures more than that of the material containing circumferential hydride.

Effect of specimen thickness on DHC velocity for Zr-2.5Nb alloy pressure tube material

Zr-2.5Nb alloy pressure tubes used in pressurized heavy water reactors (PHWR) are susceptible to failure by Delayed Hydride Cracking (DHC), which is a form of localized hydride-embrittlement phenomenon manifested in the presence of a hydrostatic stress gradient and hydrogen concentration above a threshold limit as a sub-critical crack growth process in hydride forming metals. The DHC parameters used for safety assessment are DHC velocity (VDHC) and a threshold stress intensity factor (KIH). In this work DHC velocity was determined for Zr-2.5Nb pressure tube material at 250 and 280 °C using specimens of thickness between 1 mm and 4.5 mm. The DHC velocity was found to increase with increase in specimen thickness at 250 °C and 280 °C. Significant amount of tunneling of the crack was observed for 1-mm and 2-mm thick specimens at 250 °C and 280 °C, with the degree of tunneling increasing with decrease in thickness and increase in test temperature.

Tensile properties and fracture toughness of Zr–2.5Nb alloy pressure tubes of IPHWR220

The pressure tubes of Indian Pressurized Heavy Water Reactor (IPHWR) of 220MWe are made of Zr–2.5Nb alloy manufactured either from Double Melted (DM) or from Quadruple Melted (QM) ingots. These pressure tubes are manufactured by hot extrusion, two stages of cold pilgering with intermediate annealing and autoclaving. To achieve good in-reactor performance, it is required to have minimum variability in the mechanical properties of the pressure tube across its length and between tube to tube. In this work, tensile properties and fracture toughness parameters (Jmax, dJ/da and CCL determined as per ASTM E1820-11 standard) of unirradiated Zr–2.5Nb alloy pressure tubes manufactured from DM and QM ingots using samples obtained from front and back end of the tubes is presented. The mechanical properties were evaluated in temperature range of 25–450°C and compared with the corresponding data reported in literature for CANDU pressure tubes.

Delayed hydride cracking behavior of Zr-2.5Nb alloy pressure tubes for PHWR700

In order to attain improved in-reactor performance few prototypes pressure tubes of Zr-2.5Nb alloy were manufactured by employing forging to break the cast structure and to obtain more homogeneous microstructure. Both double forging and single forging were employed. The forged material was further processed by employing hot extrusion, cold pilgering and autoclaving. A detailed characterization in terms of mechanical properties and microstructure of the prototype tubes were carried for qualifying it for intended use as pressure tubes in PHWR700 reactors. In this work, Delayed Hydride Cracking (DHC) behavior of the forged Zr-2.5Nb pressure tube material characterized in terms of DHC velocity and threshold stress intensity factor associated with DHC (KIH) was compared with that of conventionally manufactured material in the temperature range of 200–283 °C. Activation energy associated with the DHC in this alloy was found to be ∼60 kJ/mol for the forged materials.