Corrosion Of Zirconium Alloys Research Articles

Effects of Hydrogenation on the Corrosion Behavior of Zircaloy-4.

Hydrogen plays an important role in the corrosion of zirconium alloys, and the degree of influence highly depends on the alloy composition and conditions. In this work, the effects of hydrogenation on the corrosion behavior of Zircaloy-4 in water containing 3.5 ppm Li + 1000 ppm B at 360 °C/18.6 MPa were investigated. The results revealed that hydrogenation can shorten the corrosion transition time and increase the corrosion rates of Zircaloy-4. The higher corrosion rates can be ascribed to the larger stress in the oxide film of hydrogenated samples, which can accelerate the evolution of the microstructure of the oxide film. In addition, we also found that hydrogenation has little effect on the t-ZrO2 content in the oxide film and there is no direct correspondence between the t-ZrO2 content and the corrosion resistance of the Zircaloy-4.

Untangling competition between epitaxial strain and growth stress through examination of variations in local oxidation

Understanding corrosion mechanisms is of importance for reducing the global cost of corrosion. While the properties of engineering components are considered at a macroscopic scale, corrosion occurs at micro or nano scale and is influenced by local microstructural variations inherent to engineering alloys. However, studying such complex microstructures that involve multiple length scales requires a multitude of advanced experimental procedures. Here, we present a method using correlated electron microscopy techniques over a range of length scales, combined with crystallographic modelling, to provide understanding of the competing mechanisms that control the waterside corrosion of zirconium alloys. We present evidence for a competition between epitaxial strain and growth stress, which depends on the orientation of the substrate leading to local variations in oxide microstructure and thus protectiveness. This leads to the possibility of tailoring substrate crystallographic textures to promote stress driven, well-oriented protective oxides, and so to improving corrosion performance.

The accommodation of lithium in bulk ZrO2

Lithium is known to accelerate the corrosion of zirconium alloys in light water reactor conditions. Identifying the mechanism by which this occurs will allow alloying additions and alternative coolant chemistries to be proposed with the aim of improved performance. Accommodation mechanisms for Li in bulk ZrO2 were investigated using density functional theory (DFT). Defects including oxygen and zirconium vacancies along with lithium, zirconium and oxygen interstitials and several small clusters were modelled. Predicted formation energies were used to construct Brouwer diagrams. These show how competing defect species concentrations change across the monoclinic and tetragonal oxide layers. The solubility of Li into ZrO2 was determined to be very low indicating that Li solution into the bulk, under equilibrium conditions, is an unlikely cause for accelerated corrosion.

Modeling of High-Temperature Corrosion of Zirconium Alloys Using the eXtended Finite Element Method (X-FEM)

Oxidation modeling in modern nuclear fuel performance codes is currently limited by the lack of coupling with mechanics, thus preventing proper description of how high-temperature oxidation impacts mechanical properties. This is mostly due to the fact that the finite difference formalism adopted in corrosion models is incompatible with the direct coupling with mechanics in the finite element modeling employed in modern nuclear fuel performance codes. In this study, a physically based zirconium alloy corrosion model called the Coupled-Current Charge Compensation (C4) model, which was initially developed for operating temperature conditions, has been updated to include high-temperature corrosion in order to provide additional critical information (e.g., oxygen concentration profile) under loss-of-coolant accident (LOCA) conditions—information lacking in existing empirical models. The C4 model was implemented in the MOOSE finite-element framework developed at Idaho National Laboratory, enabling it to be used in the BISON nuclear fuel performance code based on the MOOSE framework. To precisely track the different interfaces at a relatively low computational cost, the eXtended Finite Element Method (X-FEM) was applied in MOOSE. The model’s results were compared to those of existing empirical models as well as metallographic analysis of high-temperature oxidized Zircaloy-4 coupons. Oxygen diffusivities in the α and β phases resulting from this comparison closely agree with those found in the literature. The C4 model implemented with X-FEM in MOOSE now has the capability to accurately predict oxide, oxygen-stabilized α, and prior β phase layer growth kinetics under isothermal exposure at high temperature (1000–1500 °C). Furthermore, in contrast with the empirical models, the C4 model accounts for the finite thickness of the fuel cladding. It can predict the oxygen concentration profile evolution through the whole cladding, enabling evaluation of the remaining ductile thickness—a crucial variable for modeling the mechanical behavior of the fuel cladding under LOCA. This implementation allows direct coupling with mechanics, at a low computing cost, using finite-element-based nuclear fuel performance codes such as BISON.

Nodular corrosion of zirconium alloys in gaseous environment containing different contaminants

The nodular corrosion susceptibility of two zirconium alloys, namely Zr-2.5Nb and Zr-2.5Nb-0.5Cu, in CO2 + O2 containing different contaminants is established. The contaminants used were ethylene, chlorine, formic acid, nitric acid, sulphuric acid and moisture. Nodules formed on both alloys in presence of each contaminant with varied extent except for sulphuric acid. The presence of ethylene resulted in nodular corrosion only in presence of radiation. The occurrence and extent of nodular corrosion is discussed in light of the “global mechanism” for nodular corrosion of zirconium alloys in gaseous environments.

Characterisation of deuterium distributions in corroded zirconium alloys using high-resolution SIMS imaging

Hydrogen diffusion through the oxide grown on Zr alloys by aqueous corrosion processes plays a critical role in determining the rate of hydrogen pickup (HPU) which can result in embrittlement of fuel cladding and limit the burnup of the nuclear fuel it encapsulates. Mapping the hydrogen/deuterium distributions in these oxide layers, especially in the barrier layer close to the metal/oxide interface, is a powerful way to understand the mechanism of both oxidation and hydrogen pickup. Here we have characterised by high-resolution SIMS analysis the deuterium distribution in oxide layers on a series of Zr alloys, including autoclave-oxidised Zircaloy-4, Zr-1Nb and Zr-2.5Nb alloys, and in-flux and out-of-flux corroded Zr-2.5Nb samples. Pre-transition Zircaloy-4 samples show a high deuterium trapping ratio in the oxide and a higher diffusion coefficient than in oxides on the Nb-containing samples. Neutron irradiation increases the deuterium diffusion coefficient, the deuterium concentration in the oxide and the pickup fraction in Zr-2.5 Nb samples. Comparative NanoSIMS and EDX/SEM analysis demonstrates that the deuterium is not preferentially trapped at second phase particles in the oxides on any of the alloys studied, but there is direct evidence for trapping at the surfaces of small oxide cracks especially in Zircaloy-4 samples. The high resolution mapping of these hot-spots in 3D can provide unique information on the mechanisms of hydrogen uptake, and suggests that the development of interconnected porosity in the oxide may be the critical rate-determining mechanism that controls HPU in the aqueous corrosion of zirconium alloys in water-cooled reactors.

Characterisation of Deuterium Distributions in Corroded Zirconium Alloys Using High-Resolution SIMS

Hydrogen diffusion through the oxide grown on Zr alloys by aqueous corrosion processes plays a critical role in determining the rate of hydrogen pickup which can result in embrittlement of fuel cladding and limit the burnup of the nuclear fuel it encapsulates. Mapping the hydrogen/deuterium distributions in these oxide layers, especially in the barrier layer close to the metal/oxide interface, is a powerful way to understand the mechanism of both oxidation and hydrogen pickup. Here we have characterised by high-resolution SIMS analysis the deuterium distribution in oxide layers on a series of Zr alloys, including autoclave-oxidised Zircaloy-4, Zr-1Nb and Zr-2.5Nb alloys, and in-flux and out-of-flux corroded Zr-2.5Nb samples. Pre-transition Zircaloy-4 samples show a high deuterium trapping ratio in the oxide and a higher diffusion coefficient than in oxides on the Nb-containing samples. Neutron irradiation increases the deuterium diffusion coefficient, the deuterium concentration in the oxide and the pickup fraction in Zr-2.5 Nb samples. Comparative NanoSIMS and EDX/SEM analysis demonstrates that the deuterium is not preferentially trapped at SPPs in the oxides on any of the alloys studied, but there is direct evidence for trapping at the surfaces of small oxide cracks especially in Zircaloy-4 samples. The high resolution mapping of these hot-spots in 3D can provide unique information on the mechanisms of hydrogen uptake, and suggests that the development of interconnected porosity in the oxide may be the critical rate-determining mechanism that controls HPU in the aqueous corrosion of zirconium alloys in water-cooled reactors.

Hydrogen pickup during oxidation in aqueous environments: The role of nano-pores and nano-pipes in zirconium oxide films

Oxidation of metals by water generates hydrogen which can enter the solid causing serious degradation of its mechanical properties and may also influence the corrosion rate. The present work focuses on hydrogen pickup during the corrosion of zirconium alloys in an aqueous environment. Transmission electron microscopy using Fresnel imaging on three different samples of oxidized Zr has been used to study the type, distribution, concentration and connectivity of nano-porosity as a function of depth through the oxide layer. Extensive interconnected nano-pipes are found in the non-protective outer part of the oxide, while in the protective barrier layer closer to the metal-oxide interface, continuous nano-pipes turn into individual nano-pores. Ab initio calculations show that molecular hydrogen is formed spontaneously by the reaction of water with oxygen vacancies in zirconium oxide. Molecular dynamics simulations reveal that these H2 molecules can diffuse rapidly through nano-pores and nano-pipes as small as 0.5 nm in the oxide layer. Calculations demonstrate that molecular hydrogen dissociates spontaneously on surfaces of suboxides found experimentally at the metal-oxide interface. Oxygen vacancies in ZrO enable the ingress and diffusion of H atoms with an energy barrier of approximately 65 kJ/mol. Further diffusion of hydrogen through oxygen-saturated α-Zr metal is fast, leading to the formation of thermodynamically stable zirconium hydrides. Thus, formation and diffusion of molecular hydrogen through nano-pores in the bulk oxide and ingress of H atoms via suboxides is a possible mechanism of hydrogen pickup in any metal or alloy covered by an oxide scale that contains nano-porosity.

Corrosion characteristics of zinc–zirconium alloy in c‐SBF and its biocompatibility in vitro/in vivo

AbstractZinc‐based metals have attracted attention as biodegradable implant materials for medical applications. New alloying systems are still being tried. In this study, three kinds of extruded zinc–zirconium alloys (R1, R2, and R3) were prepared, and two of the most important performance of corrosion characteristics and biocompatibility in vitro/in vivo were carefully examined. We found that zirconium content did not linearly influence the corrosion behavior. Zirconium improved the zinc–zirconium alloy corrosion potential, but the strengthening effect was weakened due to the ZrZn 22 phase, leading to microgalvanic corrosion. The alloy with zirconium content of 0.8% (R2) showed high corrosion resistance and minimum initial corrosion rate in the corrected simulated body fluid. R2 alloy also exhibited favorable cytocompatibility and biological histocompatibility.

A multi-technique study of “barrier layer” nano-porosity in Zr oxides during corrosion and hydrogen pickup using (S)TEM, TKD, APT and NanoSIMS

We have used (S)TEM, TKD, NanoSIMS and APT to study nano-porosity in the oxide grown on deuterated Zr-1.0Nb and Zr-2.5Nb alloys. A detailed analysis of “barrier layer” nano-porosity by TEM and TKD has revealed that the oxide grain structure is much more disorganised and the nano-porosity network better developed in the rapidly oxidising post-transition alloy. Direct observations of the trapped deuterium (D) distributions from NanoSIMS analysis also shows much more penetration of the oxide layer post-transition. APT analysis shows that there is Fe and D segregation to some of the oxide grain boundaries with occasional evidence of porosity containing trapped D and H. We conclude that interconnected porosity would offer a dominant pathway for the transport of hydrogenic species to the metal substrate during the aqueous corrosion of zirconium alloys in service.

Comprehensive investigation of the role of Nb on the oxidation kinetics of Zr-Nb alloys

Waterside corrosion of zirconium alloys is currently the life-limiting degradation mechanism of fuel elements in pressurized-water reactor (PWR) systems. Today, PWRs use Zr-Nb fuel cladding designed to limit oxidation and hydrogen pickup. However, oxidation kinetics of Zr-Nb alloys can vary substantially depending on the Nb content and distribution, even in alloys of the same composition but different heat treatments. To elucidate the role of Nb on Zr-Nb oxidation kinetics and improve the fundamental understanding of Zr-Nb corrosion, nano-beam X-ray absorption near-edge spectroscopy (XANES), transmission electron microscopy (TEM), and application of the Coupled-Current Charge-Compensation (C4) Model have been performed in concert.

Evaluation of the Corrosion State of Equipment of the First Circuit of Transport Nuclear Reactors Using the Parameters of Water Chemistry

An analysis of the results of investigation into the composition of the primary coolant of power units of nuclear-powered ice breakers Soviet Union and Arktika and the Sevmorput’ light carrier has revealed that an increase in the reactor power generation above 14 TW h leads to a sharp increase in the ammonia concentration in the coolant. An assumption was made that this effect was related with the radiation-chemical synthesis of ammonia from the nitrogen dissolved in the coolant and hydrogen resulting from enhancement of corrosion of the core structural members and fuel rod claddings made of zirconium alloys. To test the proposed hypothesis on the basis of the verified MORAVA H2 software package, a procedure was developed for calculating changes in the composition of the primary coolant in water-cooled water-moderated reactors (VVER) on exposure to reactor radiation with simultaneous metering of hydrogen at a controlled rate into the primary circuit (to simulate formation of corrosive hydrogen) during irradiation. This procedure was used to perform a computational experiment to find the qualitative regularities of the effect of “corrosive” hydrogen on the characteristics of water chemistry. The dependence of the calculated ratio of molar concentrations [NH3]/[H2] on the irradiation time has been demonstrated to linearly correlate with the measured dependencies of this ratio on the power output of similar reactor units on various nuclear-powered ships. This correlation allows for the statement on a reasonable basis that the [NH3]/[H2] ratio in the primary coolant of transport nuclear units is a fair indicator of the fact of general corrosion of zirconium alloys in the core having a structural member made of alloy E110.

Isolation of Bacterial Consortia that Induced Corrosion of Zirconium Alloys

The aim of the present study was to isolate several bacterial consortia from a soil sample and to establish if they could colonize zirconium-tin alloy, such as Zircaloy-4. Two bacterial consortia containing aerobic heterotrophic bacteria and anaerobic sulfate-reducing bacteria were isolated from a soil sample. The aerobic heterotrophic bacteria exhibited a higher capability to utilize different sole carbon sources, as compared with anaerobic sulfate-reducing bacteria. Based on a morphological, biochemical, and molecular analysis, bacterial isolates were identified as Pseudomonas putida IBBHA1, Pseudomonas aeruginosa IBBHA2, Achromobacter spanius IBBHA3, Citrobacter freundii IBBSR1, Citrobacter youngae IBBSR2, and Citrobacter braakii IBBSR3. Isolated bacterial consortia which possess distinct DNA fingerprints were able to form biofilms and colonize the surface of zirconium-tin alloy coupons, although the colonization of coupons by the aerobic heterotrophic bacteria or anaerobic sulfate-reducing bacteria alone was lower compared with that observed when the coupon was immersed in a mixture of both bacterial consortia. Coupons immersed in these bacterial consortia revealed changes in the surface characteristics, which can facilitate or accelerate zirconium-tin alloy corrosion. The accumulation of corrosion products on coupons surface was less significant when the coupons were immersed solely in aerobic heterotrophic bacteria or anaerobic sulfate-reducing bacteria, compared with that observed when the coupon was immersed in a mixture of both bacterial consortia.

Corrosion of zirconium alloys demonstrated by using impedance spectroscopy

Corrosion rate of zirconium alloys is primarily studied using weight gain method. We present a complementary procedure of a continuous and instantaneous monitoring of the corrosion rate using in-situ Electrochemical Impedance Spectroscopy (EIS), which does not disturb corrosion process in an autoclave. Two zirconium alloys, Zry-4 and Zr1Nb, were exposed in a non-active VVER primary coolant at 340 °C for almost 700 days. Measurements were aimed at passing through transition. To characterize corrosion kinetics, impedance spectra were approximated using simple models of an equivalent circuit with both the resistive and capacitive elements which described corrosion phenomena and oxide growth over the whole exposure. This way the cyclic nature of the corrosion kinetics of Zry-4 was confirmed. Oxide thickness, as a form of a cumulative corrosion rate, was obtained by combining Faraday's law and Stern-Geary approximation (FLSG). Comparing instantaneous and cumulative corrosion rate, potential inaccurate determination of transition time was demonstrated. Arrhenius interpolation was applied to correlate oxide thickness calculated based on the FLSG and weight gains. It was showed that the application of FLSG is valid.

Electro-chemo-mechanical effects of lithium incorporation in zirconium oxide

Understanding the response of functional oxides to extrinsic ion insertion is important for technological applications including electrochemical energy storage and conversion, corrosion, and electronic materials in neuromorphic computing devices. Decoupling the complicated chemical and mechanical effects of ion insertion is difficult experimentally. In this work, we assessed the effect of lithium incorporation in zirconium oxide as a model system, by performing first-principles based calculations. The chemical effect of lithium is to change the equilibria of charged defects. Lithium exists in $\mathrm{Zr}{\mathrm{O}}_{2}$ as a positively charged interstitial defect, and raises the concentration of free electrons, negatively charged oxygen interstitials, and zirconium vacancies. As a result, oxygen diffusion becomes faster by five orders of magnitude, and the total electronic conduction increases by up to five orders of magnitude in the low oxygen partial pressure regime. In the context of Zr metal oxidation, this effect accelerates oxide growth kinetics. In the context of electronic materials, it has implications for resistance modulations via ion incorporation. The mechanical effect of lithium is in changing the volume and equilibrium phase of the oxide. Lithium interstitials together with zirconium vacancies shrink the volume of the oxide matrix, release the compressive stress that is needed for stabilizing the tetragonal phase $\mathrm{Zr}{\mathrm{O}}_{2}$ at low temperature, and promote tetragonal-to-monoclinic phase transformation. By identifying these factors, we are able to mechanistically interpret experimental results in the literature for zirconium alloy corrosion in different alkali-metal hydroxide solutions. These results provide a mechanistic and quantitative understanding of lithium-accelerated corrosion of zirconium alloy, as well as, and more broadly, show the importance of considering coupled electro-chemo-mechanical effects of cation insertion in functional oxides.

The role of hydrogen in zirconium alloy corrosion

Hydrogen enters zirconium metal as a result of the corrosion process and forms hydrides when present in quantities above the solubility limit at a given temperature. Zircaloy-4 coupons of different thicknesses (0.4 mm-2.3 mm) but identical chemistry and processing were corroded in autoclave at 360 °C for various times up to 2800 days. Coupons were periodically removed and weighed to determine weight gain, which allows follow of the corrosion kinetics. Coupon thickness differences resulted in different volumetric concentrations of hydrogen, as quantified using hot vacuum extraction. The thinnest coupons, having the highest concentration of hydrogen, demonstrated acceleration in their corrosion kinetics and shorter transition times when compared to thicker coupons. Furthermore, it was seen that the post-transition corrosion rate was increased with increasing hydrogen concentration. Corrosion rates increased only after the terminal solid solubility (TSS) was exceeded for hydrogen in Zircaloy-4 at 360 °C. Therefore, it is hypothesized that the corrosion acceleration is caused by the formation of hydrides. Scanning electron microscope (SEM) examinations of fractured oxide layers demonstrate the oxide morphology changed with hydrogen content, with more equiaxed oxide grains in the high hydrogen samples than in those with lower hydrogen content. Additionally, locations of advanced oxide growth were correlated with locations of hydrides in the metal. A hypothesis is proposed to explain the accelerated corrosion due to the presence of the hydrides, namely that the metal, locally, is less able to accommodate oxide growth stresses and this leads to earlier loss of oxide protectiveness in the form of more frequent oxide kinetic transitions.

A Study of High Temperature Stability of Zirconia Surface in Pressurized Water with High Resolution X-Ray Reflectivity

Metals and metal-alloys under hydrothermal conditions undergo accelerated corrosion and degradation which is detrimental to the integrity and safety of many industrial operations. Owing to their low thermal neutron cross-section, reasonable mechanical properties, and adequate corrosion resistance in high temperature water, zirconium alloys have been used as nuclear power plant fuel cladding materials. The presence of a passive oxide layer on the zirconium surface acts as a barrier to mitigate further corrosion of the metal, which plays a key role in sustaining the integrity of cladding. However, the continuous growth of zirconia on the alloy surface limits the life span of the cladding material and therefore reduces the burn-up efficiency of the nuclear fuel and also introduces potential safety risks. Despite the factor that there exists numerous studies on the mechanisms of in-reactor corrosion of zirconium alloys, the very initial stage of the degradation mechanism still remains unclear. With development of a hydrothermal cell dedicated for in-situ synchrotron X-ray reflectivity measurement of oxide-water interface, we successfully measured the interfacial structure at yttria-stabilized zirconia, or YSZ, (111) surface for a series of temperatures ranging from room temperature up to 360 ºC in pressurized water condition of ~20 MPa. As the temperature increases, small changes in the X-ray reflectivity are observed which is presumably attributed to atomic level modification of the interfacial structure. With increasing duration time for hydrothermal reaction at higher temperatures above 340 ºC, the change in the X-ray reflectivity patterns becomes distinct and a layered structure in addition to the ambient hydration structure could be qualitatively deduced from the measured X-ray reflectivity data. Interestingly, however, when the temperature goes back to 50 ºC, the ambient hydration structure was recovered although it was not perfectly reversible but qualitatively similar. The quantitative analysis of the X-ray patterns at different temperatures and hydrothermal reaction time will provide valuable information on the interfacial hydration structure and processes from the molecular level insight.

Corrosion of Zirconium Alloys Used for Nuclear Fuel Cladding

During operation, nuclear fuel rods are immersed in the primary water, causing waterside corrosion and consequent hydrogen ingress. In this review, the mechanisms of corrosion and hydrogen pickup and the role of alloy selection in minimizing both phenomena are considered on the basis of two principal characteristics: the pretransition kinetics and the loss of oxide protectiveness at transition. In zirconium alloys, very small changes in composition or microstructure can cause significant corrosion differences so that corrosion performance is strongly alloy dependent. The alloys show different, but reproducible, subparabolic pretransition kinetics and transition thicknesses. A mechanism for oxide growth and breakup based on a detailed study of the oxide structure can explain these results. Through the use of the recently developed coupled current charge compensation model of corrosion kinetics and hydrogen pickup, the subparabolic kinetics and the hydrogen fraction can be rationalized: Hydrogen pickup increases when electron transport decreases, requiring hydrogen ingress to close the reaction.

Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

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.