Nb Pressure Tube Research Articles

Influence of Fe content on corrosion and hydrogen pick up behavior of Zr–2.5Nb pressure tube material

The effects of Fe addition in the range of 300–1250ppm in cold worked stress-relieved Zr–2.5Nb pressure tube on oxidation and hydrogen pick up behavior have been studied after 415°C steam autoclaving. Microstructure and micro-chemistry of second phase and precipitates were characterized using electron microscope. Addition of 800ppm Fe in Zr–2.5Nb alloy led to better oxidation resistance. With further addition of Fe no significant improvement of oxidation resistance was observed but hydrogen-pickup was found to increase. Zr–Nb–Fe bearing precipitates were observed in Zr–2.5Nb alloy containing 800ppm Fe. Further addition of Fe led to formation of Zr–Fe intermetallic. The chemical state of oxide has been determined by X-ray photo electron spectroscopy. Grazing Incidence X-ray Diffraction revealed that oxide in alloys with higher Fe, contained a higher fraction of tetragonal-Zirconia which is indicative of a protective oxide film and hence better oxidation resistance of the alloy.

Comparison of J-parameters of cold worked and stress relieved Zr–2.5Nb pressure tube alloy determined using load normalization and direct current potential drop technique

The fracture toughness of Zr–2.5Nb pressure tube alloy used in Indian Pressurized Heavy Water Reactor was determined in the temperature range of 30–300°C using load normalization technique (LNT) and a comparison between the crack initiation and the crack propagation toughness parameters is presented with the values obtained using direct current potential drop (DCPD) technique. The dependence of error in J-parameters determined using LNT to that obtained using DCPD on temperature and final crack extensions was also investigated. Below 200°C, the error exhibited weak dependence on final crack extensions. Above 200°C the error in initiation toughness increased significantly while the error in propagation toughness was practically unaffected.

Study of microstructure, texture and mechanical properties of Zr–2.5Nb alloy pressure tubes fabricated with different processing routes

Abstract Different fabrication trials involving the variation in three important stages of Zr–2.5Nb pressure tube were undertaken. The variations were with respect to the mode of breaking the cast structure of the ingot (forging vs extrusion), the hot extrusion ratio and the number of subsequent cold work stages to produce the finished tube. It was observed that the forging process resulted in superior performance in breaking the cast structure. Higher extrusion ratios resulted in more favorable texture and microstructure. More continuity of the beta phase was observed in the final microstructure for the route involving the single cold work step subsequent to hot extrusion.

Artificial Neural Network Modeling of In-Reactor Axial Elongation of Zr2.5%Nb Pressure Tubes at RAPS 4 PHWR

A model is developed to predict the in-reactor dimensional changes of the pressure tube materials in Indian pressurized heavy water power reactors (PHWRs) using artificial neural networks (ANNs). The inputs of the ANN are the alloy composition of the tube (concentration of Nb, O, N, and Fe), mechanical properties (yield strength, ultimate tensile strength, and percentage elongation), tube thickness, temperature, and fluence whereas axial elongation is the output. Measured elongation data from various tubes used in Indian PHWRs at Rajasthan Atomic Power Station (RAPS 4) are employed to develop the model. A three-layer feed-forward ANN is trained with the Levenberg-Marquardt training algorithm. It has been shown that the developed ANN model can efficiently and accurately predict the axial elongation of pressure tubes. The results show the high significance of Fe concentration and irradiation fluence in determining axial elongation.

Anisotropic deformation of Zr–2.5Nb pressure tube material at high temperatures

Abstract Zr–2.5Nb alloy is used for the pressure tubes in CANDU® reactor fuel channels. In reactor, the pressure tube normally operates at 300 °C and experiences a primary coolant fluid internal pressure of approximately 10 MPa. Manufacturing and processing procedures generate an anisotropic state in the pressure tube which makes the tube stronger in the hoop (transverse) direction than in the axial (longitudinal) direction. This anisotropy condition is present for temperatures less than 500 °C. During postulated accident conditions where the material temperature could reach 1000 °C, it might be assumed that the high temperature and subsequent phase change would reduce the inherent anisotropy, and thus affect the deformation behaviour (ballooning) of the pressure tube. From constant-load, rapid-temperature-ramp, uniaxial deformation tests, the deformation rate in the longitudinal direction of the tube behaves differently than the deformation rate in the transverse direction of the tube. This anisotropic mechanical behaviour appears to persist at temperatures up to 1000 °C. This paper presents the results of high-temperature deformation tests using longitudinal and transverse specimens taken from as-received Zr–2.5Nb pressure tubes. It is shown that the anisotropic deformation behaviour observed at high temperatures is largely due to the stable crystallographic texture of the α-Zr phase constituent in the material that was previously observed by neutron diffraction measurements during heating at temperatures up to 1050 °C. The deformation behaviour is also influenced by the phase transformation occurring at high temperatures during heating. The effects of texture and phase transformation on the anisotropic deformation of as-received Zr–2.5Nb pressure tube material are discussed in the context of the tube ballooning behaviour. Because of the high temperatures in postulated accident scenarios, any irradiation damage will be annealed from the pressure tube material and thus the unirradiated material results presented in this paper are also applicable to irradiated pressure tubes.

Microstructural Studies of Zr-2.5Nb and Zircaloy-2 Pressure Tubes Irradiated in Indian Pressurized Heavy Water Reactors

The Indian PHWR uses Zr-2.5% Nb pressure tubes and its in-reactor performance mainly irradiation creep and growth depends strongly on its microstructure. A detailed microstructural examination was carried out on unirradiated pressure tubes off-cuts and an irradiated pressure tube S-07 of KAPS-2 (operated for 8 effective full power years (EFPYs)), Microstructural characterization was carried out using transmission electron microscopy. Microstructual observation of un-irradiated off-cuts shows the lamellar morphology of the -Zr along with the -phase present as stingers between two alpha laths as well as fine and coarse beta globules. The size of -Zr lamellae was found to be in the range from 0.17 to 0.2 m, 1.8 to 2.4 m and 1.7 to 2.8 m in the radial, circumferential and axial direction respectively (aspect ratio of 1:7:8). TEM-EDS analysis showed composition of the  phase tin the range of 15-50 wt%Nb. The irradiated pressure tube samples obtained from 13 locations were showing average alpha grain width, grain length and aspect ratio in the range of 0.17-0.27 micron, 1.7-2.3 micron and 7.1-8.5 respectively. Extensive modification in beta morphology could be seen at the high flux and high temperature regions. The  phase was observed to have globulised completely in many regions. They were present at the interface of -Zr laths as well as within the lath. The Nb concentration of the  phase appeared to have increased as the volume fraction had reduced. The microstructure details of irradiated and un-irradiated pressure tubes obtained in this study is expected to help in modeling the dimensional change occurring during irradiation in reactor.

High temperature steam oxidation study on Zr–2.5%Nb pressure tube under simulated LOCA condition

Study of high temperature steam oxidation kinetics and microstructural evolution during the oxidation was carried out on the coupons of Zr–2.5%Nb pressure tube material of Indian Pressurized Heavy Water Reactors (PHWRs) in the temperature range 500–1050°C. The oxidation kinetics derived from the weight gain measurements showed a parabolic rate law with the parabolic rate constant KP expressed as an Arrhenius equation KP=10.12×108×exp(−18664/T). Hydrogen pick up was less than 55ppm in the samples oxidized at temperatures up to 850°C but high (250–400ppm) in the samples oxidized in the β phase region (900°C and above). The microstructure of the samples oxidized above the α-Zr/β-Zr transition temperature showed from the surface inwards sequentially the presence of an oxide layer, an underlying oxygen stabilized α-Zr layer and a prior β-Zr phase containing hydride precipitates. An increase in the hardness was observed near the oxide-metal interface in the coupons oxidized above 900°C, due to formation of oxygen stabilized α-Zr layer. Higher hardness was also observed in the base metal in the samples oxidized at 1000°C and 1050°C.

Hydride reorientation in Zr2.5Nb studied by synchrotron X-ray diffraction

The crystallographic texture and stress state of hydride platelets of two different orientations precipitated in Zr2.5Nb pressure tubes were studied by synchrotron X-ray diffraction experiments using an 80keV photon beam and an area detector in transmission geometry. Circumferential and radial hydrides with platelet normals along the tube radial and circumferential directions, respectively, were precipitated in tube material loaded up to H contents of 130wt.ppm. This macroscopic hydride orientation was controlled by application of a load along the tube hoop direction during hydride precipitation. The experiments show that both circumferential and radial platelets are composed of δ-hydrides precipitated in α-Zr grains from a wide range of orientations, but with a clear preference for Zr crystals with their c-axes at an angle of ∼15–20° from the hoop direction. Some moderate differences between the crystallographic texture of the two hydrides result from application of the load during precipitation. The results are explained in terms of an autocatalytic nucleation process and the Zr2.5Nb microstructure. A careful stress analysis revealed that both hydride types are compressed by the matrix on the plane of the platelet, with the largest stresses always found along the axial direction of the tube. Determination of the stress state of the hydride could be exploited as a diffraction signature of the hydride orientation.

Influence of hydrogen content on fracture toughness of CWSR Zr–2.5Nb pressure tube alloy

In this work, influence of hydrogen and temperature on the fracture toughness parameters of unirradiated, cold worked and stress relieved (CWSR) Zr–2.5Nb pressure tube alloys used in Indian Pressurized Heavy Water Reactor is reported. The fracture toughness tests were carried out using 17mm width curved compact tension specimens machined from gaseously hydrogen charged tube-sections. Metallography of the samples revealed that hydrides were predominantly oriented along axial–circumferential plane of the tube. Fracture toughness tests were carried out in the temperature range of 30–300°C as per ASTM standard E-1820-06, with the crack length measured using direct current potential drop (DCPD) technique. The fracture toughness parameters (JQ, JMax and dJ/da), were determined. The critical crack length (CCL) for catastrophic failure was determined using a numerical method. It was observed that for a given test temperature, the fracture toughness parameters representing crack initiation (JQ) and crack propagation (JMax, and dJ/da) is practically unaffected by hydrogen content. Also, for given hydrogen content, all the aforementioned fracture toughness parameters increased with temperature to a saturation value.

Deducing the stress–strain response of anisotropic Zr–2.5%Nb pressure tubing by spherical indentation testing

In this paper we developed a method of analysis by which the average stress–plastic strain flow curve of mechanically anisotropic materials can be deduced from spherical indentation test data. Our analysis is based upon spherical indentation tests performed on the extruded and cold-drawn Zr–2.5%Nb CANDU pressure tube material over the range of temperature from 25°C to 300°C. The indentation force and depth data were analyzed and σavg and ɛavg were calculated using previously reported equations developed for spherical indentation of isotropic material which were then modified, by incorporating the appropriate Hill’s anisotropy coefficients, to characterize the anisotropic yield stress of the indented material. The resulting flow curves were dependent on indentation direction and correspond closely with flow curves obtained from previously reported conventional uniaxial stress tests performed on the Zr–2.5%Nb material. Indentation tests performed with large, 200μm, and small, 40μm, diameter spheres indicate that for small diameter indentations, when the indentation depth is less than several micrometers, the calculated σavg is heavily influenced by the depth dependence of the yield strength of the indented material.

Corrosion of zirconium alloys in concentrated lithium hydroxide solutions

The accelerated corrosion of two alloys used as fuel cladding material, such as Zircaloy-4 and Zr–1%Nb has been studied in concentrated lithium hydroxide (LiOH) solutions at high temperature and pressure. Zr–2.5%Nb pressure tube material (PT) was also tested in order to analyze the influence of the amount of β-Zr phase on the accelerated corrosion of the Zr–Nb alloys. The microstructure of Zircaloy-4 consisted of α-Zr equiaxed grains whereas that of Zr–2.5%Nb (PT) and Zr-1%Nb alloys showed a two phase (α-Zr+β-Zr) microstructure.Autoclaving tests were performed in LiOH solutions with concentrations ranging from 0.1M to 1M for 16h at 343°C, and also in steam at 400°C. The extent of corrosion was evaluated through the weight gain and visual appearance of the oxide. Measurements included hydrogen uptake and oxide thickness. Optical microscopy observations of the hydride distribution were also made on the specimens.Results showed that Zircaloy-4 suffered accelerated corrosion at 0.7M and 1MLiOH concentrations with high hydrogen uptakes (∼50%). For these LiOH concentrations, although the Zr–2.5%Nb (PT) and Zr–1%Nb alloys showed weight gains higher than the threshold value established for the “transition”, their hydrogen uptakes were low (∼1%). This behavior indicates that at the early stages of the corrosion process in these solutions, an oxide barrier layer may be present at the metal/oxide interface of these Zr–Nb alloys. The lower amount of β-Zr phase in Zr–1%Nb improves the characteristics of the oxide layer in relation to that formed on Zr–2.5%Nb and does not affect the hydrogen uptake.

Assessing Long-Term Performance of CANDU®Out-of-Core Materials

As so-called second-generation power reactors are approaching the end of their original design lives, assessments are being made to determine the feasibility and economics of extending plant life. Although components exposed to neutron and gamma irradiation are often those of most concern in terms of in-service ageing and continued fitness for service, ageing of out-of-core components can also limit the possibility of extended service life beyond design life. In CANDU®reactors, life extension decisions occur when the Zr-2.5Nb pressure tubes reach end of life, typically after about 25 years of service for the first CANDU-6 units. At the time of pressure tube replacements, the remaining life predictions for several other major components or systems provide the information required to determine life extension feasibility. Several CANDU reactors are currently being refurbished, with others planned, and experience to date shows that the steam generators, heat transport system piping and various balance of plant piping systems are typically those requiring careful assessment to ensure successful refurbishment. In this paper, we discuss AECL R&D that is oriented towards providing the chemistry and materials inputs required to assess current condition and predict future ageing of CANDU reactor out-of-core components and systems, and in particular steam generators (Alloy 800 tubing and carbon steel internals), feeder pipes and related heat transport system piping (carbon steel flow accelerated corrosion, feeder cracking). Systems and components that may impact future life will also be discussed, along with the related R&D, and this includes balance of plant system piping (feedwater piping and buried piping), cables and concrete structures.

Assessment of the anisotropic flow stress and plastic strain of Zr–2.5%Nb pressure tubes at temperature from 25 °C to 300 °C

In this paper uniaxial compression testing is used to assess the stress versus plastic strain response in the axial, radial and transverse directions of the mechanically anisotropic Zr–2.5%Nb CANDU pressure tube material over the temperature range from 25 to 300°C. The data from these tests are used to determine the Hill’s anisotropy coefficients F=0.38±0.01, G=0.202±0.001, and H=0.62±0.01 of the pressure tube material. These coefficients were found to be independent of temperature and plastic strain. The results of this study represent the only experimentation-based determination of the Hill’s anisotropy coefficients F, G, and H of extruded and cold-drawn Zr–2.5%Nb CANDU pressure tubes over a wide temperature range extending up to the 300°C in-service temperature of these pressure tubes.

A Differential Scanning Calorimetry (DSC) Study of Phase Changes in an As-Received Zr-2.5Nb Pressure Tube Material during Continuous Heating and Cooling

Differential scanning calorimetry (DSC) has been used to study the phase changes in samples of as-received Zr-2.5Nb pressure tube material by continuous heating and cooling. Two different heating rates (5 and 20°C/min) were used to heat the sample up to 1050°C. After a short time hold at 1050°C, all the samples were continuously cooled to 300°C at a rate of 20°C/min. On continuous heating, the DSC signals obtained showed two endothermic transitions. The low-temperature transition, occurring between about 500 and 650°C, is attributed to a thermal decomposition of metastable niobium-stabilized β-phase. The highertemperature transition, occurring between 600 and 950°C, is due to phase transformations of hcp α-Zr to bcc β-Zr, as previously confirmed in a companion study on the same pressure-tube material that was examined in-situ by neutron diffraction. The neutron diffraction results provided a positive identification of the two phases and also a quantification of the β-phase present in the sample at different heating temperatures, and thus provided a guide to extract the volume fraction of β-phase from the DSC signals obtained in this study. The DSC signals revealed only one exothermic transition which is correlated to the reverse transformation of β-Zr to α-Zr, as previously identified in the companion neutron diffraction study of the same pressure tube material.

Anisotropic deformation of Zr ion irradiated Zr–2.5%Nb micro-pillars at 25 °C

Micro-pillars of 5 μm diameter and 5 μm height were made from textured Zr–2.5%Nb pressure tube material and were irradiated with 8.5 MeV Zr + to simulate neutron irradiation. The yield stress, strain hardening exponent, and degree of strain localization of the micro-pillars when tested at 25 °C increased with ion irradiation and the extent of increase was largest in directions containing low (0 0 0 1) basal pole fraction. This suggests that Zr + irradiation inhibits prismatic dislocation slip more than pyramidal slip in this material.

Crystallographic Texture and Volume Fraction of α and β Phases in Zr-2.5Nb Pressure Tube Material During Heating and Cooling

The phase transformations in an as-received Zr-2.5Nb pressure tube material were characterized in detail by neutron diffraction. The texture and volume fraction of α and β phases were measured on heating at eight different temperatures 373 K to 1323 K (100 °C to 1050 °C) traversing across the α/(α + β) and (α + β)/β solvus lines, and also upon cooling at 1173 K and 823 K (900 °C and 550 °C). The results indicate that the α-phase texture is quite stable, with little change in the {0002} and \( \left\{ {11\bar{2}0} \right\} \) pole figures during heating to 1123 K (850 °C). The β-phase volume fraction increased while a slight change in texture was observed until heating reached 973 K (700 °C). On further heating to 1173 K (900 °C), there appears a previously unobserved α-phase texture component due to coarsening of the prior primary α grains; meanwhile the transformed β-phase texture evolved markedly. At 1323 K (1050 °C), the α phase disappeared with only 100 pct β phase remaining but with a different texture than that observed at lower temperatures. On cooling from the full β-phase regime, a different cooldown transformed α-phase texture was observed, with no resemblance of the original texture observed at 373 K (100 °C). The transformed α-phase texture shows that the {0002} plane normals are within the radial-longitudinal plane of the pressure tube following the Burgers orientation relationship of (110)bcc//(0002)hcp and \( [\bar{1}11]_{\text{bcc}} //[11\bar{2}0]_{\text{hcp}} \) with a memory of the precursor texture of the primary α grains observed on heating at 1173 K (900 °C).

Microstructural and Textural Evolution in Heat Treated Zr-2.5% Nb Pressure Tube Material Subjected to Dilatometric Studies

β-transus boundary for the Zr-2.5%Nb alloy containing 1137 ppmw of O and 650 ppmw of Fe was established using quenching dilatometry technique. The solid cylindrical samples of the alloy were β homogenised at 1063°C and then quenched at different rates between 0.06 and 200°C/s using the quenching dilatometer. The cooling part of the dilatograms were analysed to establish continuous cooling β transus boundary. Based on this β transus boundary heat treatment experiments were carried out on the alloy samples by holding at temperatures 883, 870, 863, 840°C in the α + βZr phase field and then quenching at different rates between 0.5 and 100°C/s to study the microstructural evolution and textural changes using electron microscopy techniques. An attempt has been made to correlate the transformation characteristics with the dilatograms recorded and the microstructural and textural changes have been ascertained.

Temperature dependence of the anisotropic deformation of Zr–2.5%Nb pressure tube material during micro-indentation

The effect of temperature on the anisotropic plastic deformation of textured Zr–2.5%Nb pressure tube material was studied using micro-indentation tests performed in the axial, radial, and transverse directions of the tube over the temperature range from 25 to 400 °C. The ratio of the indentation stress in the transverse direction relative to that in the radial and axial directions was 1.29:1 and 1.26:1 at 25 °C but decreased to 1.22:1 and 1.05:1 at 400 °C. The average activation energy of the obstacles that limit the rate of indentation creep increases, from 0.72 to 1.33 eV, with increasing temperature from 25 to 300 °C and is independent of indentation direction. At temperature between 300 °C and 400 °C the measured activation energy is considerably reduced for indentation creep in the transverse direction relative to that of either the axial or radial directions. We conclude that, over this temperature range, the strength of the obstacles that limit the time-dependent dislocation glide on the pyramidal slip system changes relative to that on the prismatic slip system. These findings provide new data on the temperature dependence of the yield stress and creep rate, particularly in the radial direction, of Zr–2.5%Nb pressure tubes and shed new light on the effect of temperature on the operation of dislocation glide on the prismatic and pyramidal slip systems which ultimately determines the degree of mechanical anisotropy in the highly textured Zr–2.5Nb pressure tube material used in CANDU nuclear reactors.

Comparison of the measured and predicted crack propagation behaviour of Zr–2.5Nb pressure tube material

The crack propagation behaviour of Zr–2.5Nb pressure tube material was studied through comparison of the measured and predicted behaviour. Three-dimensional finite element simulations of compact tension and burst specimens were performed using the crack tip opening angle (CTOA) as a fracture criterion to allow for crack propagation. To obtain reasonable agreement with the measured force, crack extension, displacement, and crack tunnelling behaviour from compact tension simulations, a non-constant CTOA profile was required. This CTOA profile was then used as the crack propagation criterion in simulations of burst specimens where it was again found that there was reasonable agreement obtained with experiment.

Improved Zr-2.5Nb Pressure Tubes for Reduced Diametral Strain in Advanced CANDU Reactors

Abstract In an Advanced CANDU Reactor (ACR) (ACR is a registered trademark of Atomic Energy of Canada Limited), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure, and flux than those in a CANDU (CANDU is a registered trademark of Atomic Energy of Canada Limited) reactor, and accordingly require a thicker wall (6.5 mm for ACR versus 4.2 mm for CANDU). In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a study to model and forecast the performance of these thicker pressure tubes has been undertaken. One of the main requirements for the pressure tube is to have low diametral creep. Based on previous experience with CANDU reactor pressure tube performance and manufacture, an assessment of the grain structure and texture of the ACR pressure tubes indicates that the in-reactor creep deformation will be improved. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor.