Solubility Limit Of Hydrogen Research Articles

The role of symmetric tilt grain boundaries on the precipitation of hydrides in zirconium: A molecular dynamics study

Zirconium (Zr) alloys are widely used as structural materials in the core of nuclear reactors. However, the absorption of hydrogen by these alloys during the reactor operation may lead to the precipitation of hydrides. Such hydrides have considerable effects on the microstructure and mechanical properties of the alloys, thus significantly impacting their overall performance. In this work, we investigate the behavior of hydrogen in the presence of symmetric tilt grain boundaries in Zr using molecular dynamics simulations. To systematically explore the conditions for the formation of hydrides, we consider hydrogen concentration levels of up to 50 at.%. We find that hydrogen concentrations of 30 at.% or lower are below the threshold for the precipitation of hydrides, while fcc coordinated hydrides precipitate at 50 at.%, which is very close to the solubility limit of hydrogen in Zr at 800 K. We find that the grain boundaries (GBs) play a pivotal role in the behavior of hydrogen, as well as the formation of hydrides in Zr systems. In most cases, large amounts of hydrogen are found to accumulate at GBs. However, both homogeneous and heterogeneous nucleation are observed during the formation of hydrides at the grain interiors and GBs, respectively. Nonetheless, in some cases the GB inhibit the formation of hydrides.

Irradiation growth behavior and effect of hydrogen absorption of Zr-based cladding alloys for PWR

In order to understand the dimensional stability of the fuel rod during long-term use in commercial LWRs, irradiation growth testing on various Zr-based fuel cladding materials including improved Zr-based alloys was conducted in the Halden reactor in Norway under prototypic irradiation and corrosion environment. The effect of hydrogen absorbed in the cladding tube due to corrosion on the irradiation growth behavior was investigated in this paper. The comparison of specimens with or without pre-charged hydrogen (about 200 and 400 ppm) revealed that the effect of hydrogen absorption, i.e. the acceleration of irradiation growth, became significant when the hydrogen content exceeded the hydrogen solubility limit in the corresponding irradiation temperature. Based on this understanding, the growth acceleration effect was evaluated to be (0.06 ± 0.01)%/100 ppm, with the denominator defined as the amount of absorbed hydrogen involved in hydride precipitation under irradiation as a relevant parameter. Discrimination of hydrogen effect confirmed lower irradiation growth of the improved Zr-based alloys than the conventional Zry-4, which may be attributed to the effect of lower hydrogen absorption property for Nb addition specimens.

Hydrogen absorption and hydride formation in pure titanium T40 (grade 2) and TA6V ELI (grade 23) under cathodic polarization in artificial seawater

Different kinetics of hydrogen absorption in T40 (grade 2) and TA6V ELI (grade 23) under cathodic polarization in artificial seawater have been highlighted. These polarizations were made by applying potentials from -0.8 to -1.8V/SCE in artificial seawater and NaCl solution. Four stages were identified and related in term of hydrogen ingress, hydrides formation and calcareous deposit growth. The formation of γ and δ-hydrides have been observed, localized and characterized using several techniques. On T40, hydrides form as a layer that increases the surface roughness and clusters form in the bulk after first moments of hydrogen absorption. Whereas in TA6V ELI, hydrogen is absorbed by β-phase leading to a volume expansion of this phase. Then after reaching the hydrogen solubility limit of β-phase, hydrides form on interfaces α/β or α/α and in α grains. For long durations, the hydrogen ingress is limited by the subsurface hydrides and the stabilized calcareous deposit. These different steps are time depend on processes which need to be tacked into account to improve knowledge of hydrogen embrittlement in titanium alloys.

Implementation of Hydrogen Solid Solubility Data and Precipitation Threshold Stresses in the Fuel Rod Code TESPA-ROD

During operation of light water reactors, the Zircaloy fuel rod cladding is susceptible for hydrogen uptake. When the local solubility limit of hydrogen in Zircaloy is reached, additional hydrogen precipitates as zirconium hydride, which affects the ductility of the fuel rod cladding. Especially, the radially aligned hydrides enhance embrittlement, while circumferential (azimuthal) hydrides have a less detrimental effect. In this work, the influence of high temperatures during the dry storage period on hydride dissolution and precipitation is demonstrated. Therefore, in a descriptive example scenario being discussed, the simulation of a limited heat removal from the cask will heat up the dry storage cask for days and causes dissolution of hydrides in the cladding. Depending on the threshold stress for reorientation, the following cooldown results on different hydride precipitation behavior. The threshold stress leads to an enhanced or delayed precipitation of radial hydrides. The GRS fuel rod code TESPA-ROD is equipped with a new model for hydrogen solubility and applied to long-term storage transients. In this article, hydride refers to zirconium hydrides formed inside the fuel rod cladding.

Direct measurement of hydrogen diffusivity and solubility limits in Zircaloy 2 (formula unit of ZrH0.0155) using incoherent quasi-elastic neutron scattering

The diffusivity of hydrogen is an important property of light water nuclear reactor (LWR) fuel cladding. LWR cladding absorbs hydrogen during normal operation, a contributing factor to embrittlement that decreases the lifetime of the fuel. Mass transport of hydrogen is dictated by an Arrhenius behavior typical of solid state diffusion and the associated activation energy is therefore a property relevant to LWR fuel performance. We have used incoherent quasi-elastic neutron scattering (QENS) to directly measure the diffusivity of hydrogen in recrystallized Zircaloy 2 with a hydrogen concentration of 170 μg/g. We rely upon the low-Q expansion for long-range diffusion to determine diffusivity as a function of temperature between 572 and 780 K. We find the diffusivity is given by D(T) = 0.0067 exp (-0.461 eV/kT) [cm2/s] below 670 K and by D(T) = 0.0012 exp (-0.36 eV/kT) [cm2/s] above 670 K. Our activation energy below 670 K agrees with the value typically used to assess hydrogen diffusivity in LWR cladding [Kearns, Journal of Nuclear Materials 43 (1972) 330], but is approximately 20% lower above 670 K. The two different activation barriers are attributed to impurity trapping of hydrogen solutes at lower temperature that ceases to influence diffusivity at higher temperature. The application of the Oriani model for diffusion with impurity trapping to our system demonstrates the plausibility of this hypothesis. We believe this mechanism may be responsible for historical discrepancies of measured hydrogen diffusivity in Zr-based alloys. The elastic intensity versus temperature in fixed window scans exhibit inflection points that are in good agreement with the published terminal solid solution solubility limits for hydrogen in Zircaloy 2.

Study hydrogen embrittlement and determination of E110 fuel cladding mechanical properties by ring compression testing

.Zirconium based alloys are commonly used as material for fuel claddings in the light water reactors. Claddings act as first metallic barriers against loss of fission products during the nuclear power plant operation, intermittent storage or final dry storage. The integrity of claddings is always critical issue during reactor operation and during storage of spent fuel. In this work, ring compression testing method developed was applied to study hydrogen embrittlement, to evaluate the stress-strain behaviour and hoop fracture properties of E110 (Zr-based) fuel claddings. The results show that the collapse load and the tensile strength values depend strongly on hydrogen concentration. In particular, tensile strength experiment data shows significant change in its trend after reaching the maximum hydrogen solubility limit.

Imaging and hydrogen analysis by SIMS in zirconium alloy cladding: a dual ion beam approach

Zirconium‐based alloys are used as fuel cladding material in nuclear power plants. During the reactor operation, the material is oxidized by the water (reactor coolant), and hydrogen gas is formed. The released hydrogen gas, which partly diffuses into the zirconium alloy, forms hydrides when the hydrogen solubility limit is exceeded. This may result in blistering and cracking of the cladding – a phenomenon commonly known as hydrogen embrittlement. Both corrosion and H‐uptake phenomena are limiting factors, which can reduce the time of life of cladding under irradiation in a power plant. This study concerns a preliminary investigation of hydrogen by SIMS in a hydrogenated but not‐irradiated Zircaloy‐4 sample in order to develop a better understanding of hydrogen ingression in high burn up irradiated cladding tubes. Although hydrogen can be easily detected using the SIMS technique, any contamination source of hydrogen must be well identified and minimized during the SIMS measurements. In this work, we have analyzed a hydrogenated (135 wppm in bulk) Zircaloy‐4 cladding specimen. In the first step, the sample's surface has been presputtered with cesium ions. Thereafter, H− ions yield images have been acquired analyzing the sample with a gallium primary ion beam. Using this dual beam approach, the hydride precipitates present in the sample have been clearly identified and analyzed. Copyright © 2014 John Wiley & Sons, Ltd.

Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick’s law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

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.

Comparison of delayed hydride cracking behavior of two zirconium alloys

Delayed hydride cracking (DHC) is an important failure mechanism that may occur in Zr alloys during service in water-cooled reactors. Two conditions must be attained to initiate DHC from a crack: the stress intensity factor must be higher than a threshold value called KIH and, hydrogen concentration must exceed a critical value. Currently the pressure tubes for CANDU reactor are fabricated from Zr–2.5Nb. In this paper the critical hydrogen concentration for DHC and the crack velocity of a developmental pressure tube, Excel, was evaluated and compared with that of Zr–2.5Nb. The DHC velocity values measured in Excel were higher than usually reported in Zr–2.5Nb. Due to the higher hydrogen solubility limits in Excel, its critical hydrogen concentration for DHC initiation is 10–50wppm over that of Zr–2.5Nb in the range of 150–300°C.

Identification and characterization of a new zirconium hydride

Zirconium alloys are currently used in nuclear power plants where they are susceptible to hydrogen pick-up. Hydride precipitation may occur when the hydrogen solubility limit is reached. Various Zr hydride phases, gamma, delta and epsilon have been identified since the 1950s. Combining electron precession microdiffraction, electron energy loss spectroscopy and ab initio electronic calculations, a new Zr hydride named zeta has been identified and characterized. It belongs to the trigonal crystal system with space group P3 m1 and it is fully coherent with the alphaZr matrix.

Molecular Hydrogen in Amorphous Silicon with High Internal Stress

Hydrogenated amorphous silicon with high compressive stress and hydrogen concentration as high as the hydrogen solubility limit has been studied. The concentrations of total hydrogen and bonded hydrogen have been determined by nuclear reaction analysis and infrared transmission spectroscopy, respectively. The amount of molecular hydrogen has been estimated as a difference between these two concentrations. Silicon ion implantation has been used to transform the hydrogen molecules into silicon-bonded hydrogen. The sensitivity of the Fourier transform infrared spectrometer to the expected variations in the bonded hydrogen concentration has been proved using hydrogen implantation. It has been concluded that in spite of high hydrogen content in the material the concentration of molecular hydrogen in the studied films is below 1 at. % and is not the reason for the high internal stress.

Hydrogen solubility limit in hydrogenated amorphous silicon

Hydrogen solubility has been studied in hydrogenated amorphous silicon (a-Si:H) prepared by plasma-enhanced chemical vapour deposition with hydrogen-diluted silane. Post-hydrogenation experiments have been carried out using hydrogen plasma and hydrogen ion implantation. Thermal annealing and silicon ion implantation have been used to change the defect density in the amorphous silicon network. Hydrogen concentration has been established by means of nuclear reaction analysis and infrared spectroscopy. It has been shown that the hydrogen solubility in a-Si:H is strongly related to the density of the hydrogen trapping sites in the silicon network and the value of the solubility limit is determined by the material structure and consequently by the a-Si:H preparation conditions. The ratio between the hydrogen concentration and its solubility limit has been discussed in the context of the light-induced degradation of a-Si:H.

Laser hydriding of crystalline and amorphous silicon

Crystalline and ion-beam-amorphized silicon samples were irradiated with a pulsed nanosecond excimer laser in a pure hydrogen atmosphere. The hydrogen concentration was determined via the 1H(15N,αγ)12C nuclear reaction. In the case of crystalline silicon, hydrogen incorporation into the sample surface was found to be well above the hydrogen solubility limit at thermodynamic equilibrium. The hydrogen depth profiles perfectly matched the damage profiles measured via Rutherford backscattering channeling spectroscopy. For the pre-amorphized silicon samples, the laser treatment resulted in epitaxial recrystallization of the amorphous top layer, but the hydrogen uptake was found to be negligible in that case. The experimental data were compared with the results of thermodynamic simulations of the laser–gas–material interaction.

Hot-wire deposition of photonic-grade amorphous silicon

A new amorphous silicon application related to the patterning of refractive index for the purpose of defining and integrating photonic-device elements is emerging. Photonic device elements include waveguides, splitters, mirrors, optical memories, etc. Hot-wire-deposited amorphous silicon has several attributes that make it an exceedingly attractive matrix for photonic device patterning, including: high hydrogen solubility limits; relatively little sub-gap absorption; low stress; non-peeling films; and fast, economical deposition of thick (≥5 μm) films, as well as optically smooth as-deposited surfaces, even on thick films. The growing catalog of proposed and/or demonstrated amorphous silicon-based optical devices is rapidly expanding.

A neutron-spectroscopy study of the local vibrations, the interstitial sites and the solubility limit of hydrogen in niobium–molybdenum alloys

We studied by neutron spectroscopy, and for temperatures from 10 to 300 K, the local vibrations of hydrogen interstitials in two niobium-based alloys containing 5 and 20 at.% molybdenum (Nb 0.95Mo 0.05H 0.03 and Nb 0.8Mo 0.2H 0.05). Within experimental accuracy, the vibrational energies agree with those of hydrogen in pure niobium. This indicates that the hydrogen in the two alloys occupies tetrahedral interstitial sites. At 10 K, the values of the vibrational energies show a complete hydrogen precipitation in the Nb 0.95Mo 0.05H 0.03 alloy and no hydrogen precipitation in the Nb 0.8Mo 0.2H 0.05 alloy. The fact that the hydrogen in the latter alloy was still in solid solution means that the molybdenum alloy components lead to an increase of the solubility limit of the hydrogen.

A theoretical estimation of fracture toughness of zirconium hydride

The presence of hydrogen in zirconium alloys causes the hydrides to precipitate when hydrogen solubility limits are exceeded. Hydrides in zirconium alloys can lead to a significant embrittlment effect under certain circumstances, and mechanical properties, especially crack growth resistance, may be affected. As a result, there is a phenomenon of delayed hydride cracking (DHC) of zirconium alloys which is characterized by crack propagation at low values of the stress intensity factor. One of important mechanical properties of zirconium alloys is the threshold stress intensity factor, KIH, below which cracks do not grow. Therefore, the problem of hydrogen embrittlment of zirconium alloys is a very important, particularly for structural components in the nuclear industry. For example, the gradual build-up of deuterium in the Zr-2.5Nb pressure tubes of a pressurized heavy water reactor as a result of corrosion, combined with the possibility of the generation of flaws of sufficient size, makes the tubes susceptible to the initiation of DHC [1]. In this case, the service life of a pressure tube is determined by its capability to resist the crack initiation and propagation. Because of the technological importance of this phenomenon, the crack initiation threshold due to the hydride formation has been studied. A theoretical prediction of the KIH value of zirconium alloys can be based on modelling of the process zone ahead of the crack tip (e.g. [2–4]). The process zone is treated as a zone of fracture process. On the other hand, hydrided zirconium alloy can be interpreted as zirconium-zirconium hydride mixture. In this case, to predict the value of KIH of hydrided zirconium alloy it is necessary to estimate the fracture toughness of the hydride. However, measurements of fracture toughness of δ-zirconium hydrides showed large scatter due to specimen problems because of their extreme brittleness [5]. Therefore, a theoretical estimation of the value of K H IC for bydride is very important for an analysis and optimisation of crack growth resistance in zirconium alloys. This paper presents a theoretical estimation of fracture toughness of zirconium hydrides. The theory is based on a local approach to fracture of zirconium alloys. Recently, a model for the threshold stress intensity factor, KIH, was suggested leading to an analysis of crack growth initiation in zirconium alloys during DHC [4]. Local criterion of brittle fracture and modelling of the fracture process zone were used to describe fracture initiation at the hydride platelet in the process zone ahead of the crack tip. The hydride platelet was assumed to cover the process zone and the crack growth proceeds through fracturing of the hydride. The critical condition for fracture initiation in the hydride is reached when the maximum of the local stress approaches the fracture strength of the hydride at some characteristic (critical) distance, rc, ahead of the crack tip (Fig. 1). It was shown that the theoretical KIH-estimation applied to the case of mixed plane condition within the process zone is qualitatively consistent with experimental data for unirradiated Zr-2.5Nb alloy. In the framework of the proposed model, the theoretical value of KIH for a single hydride platelet at the crack tip characterizes the ultimate lower bound value of crack resistance in hydrided zirconium alloy during DHC, whereas the minimum of the threshold stress intensity factor, (KIH)min, should be considered as the minimal capability of the hydride to resist fracture within the hydride, i.e. the fracture toughness, K H IC, of hydride. The minimum fracture toughness of structural materials can be described by a model called the Kμ-model [6]. The following equation should be valid

Hydrogen Absorption and Desorption by Quasicrystalline and Related Approximant Powders Produced by Mechanical Alloying in the Mg–Al–Zn System

Icosahedral quasicrystalline (i-phase) Mg 44 Al 15 Zn 41 and related Frank-Kasper (cubic approximant) phase powders, which were produced by mechanical alloying from the elemental powder mixture and subsequent heat treatment, respectively, have been hydrogenated with hydrogen gas at a temperature of 573 K and a hydrogen pressure of 4.2 MPa. The hydrogen solubility limit for both powders is almost the same and was estimated to be about 16 at%, and no crystallographic structural dependence was found for the hydrogen solubility, The approximant phase powder was structurally stable after hydrogenation but the i-phase powder decomposed into the approximant and MgZn 2 intermetallic phases. Hydrogen desorption occurred at temperatures above about 600 K for both powders and was accompanied by the evaporation of zinc. The kinetics of the hydrogen desorption as well as the zinc evaporation were slightly different in each powder because of the formation of the MgZn 2 phase in the i-phase powder after hydrogenation.

Solubility and partitioning of hydrogen in metastable Zr-based alloys used in the nuclear industry

Abstract Terminal solubility and partitioning of hydrogen in Zr–Nb alloys with different niobium concentrations were examined using differential scanning calorimetry and hot vacuum extraction mass spectrometry. Specimens were charged to different concentrations of hydrogen and annealed at 1123 K to generate a two-phase structure consisting of α-Zr (Zr–0.6 wt.% Nb) and metastable β-Zr (Zr–20 wt.% Nb) within the alloy. Specimens were aged at 673 and 773 K for up to 1000 h to evaluate the effect of the decomposition of the metastable β-Zr to α-Zr+β-Nb on the solubility limit. The results show that the solubility limit for hydrogen in the annealed Zr–Nb alloys is higher than in unalloyed Zr and that the solubility limit increases with the niobium concentration of the alloy. They also show that the hydrogen solubility limits of the completely aged Zr–Nb alloys are similar and approach the values for pure α-Zr. The solubility ratio of hydrogen in β-Zr (Zr–20 wt.% Nb) to that in α-Zr (Zr–0.6 wt.% Nb) was found to range from 9 to 7 within the temperature range 520–580 K.

Fatigue crack growth behavior in niobium-hydrogen alloys

Near-threshold fatigue crack growth behavior has been investigated in niobium-hydrogen alloys. Compact tension specimens (CTS) with three hydrogen conditions are used: hydrogen-free, hydrogen in solid solution, and hydride alloy. The specimens are fatigued at a temperature of 296 K and load ratios of 0.05, 0.4, and 0.75. The results at load ratios of 0.05 and 0.4 show that the threshold stress intensity range (ΔK th ) decreases as hydrogen is added to niobium. It reaches a minimum at the critical hydrogen concentration (C cr ), where maximum embrittlement occurs. The critical hydrogen concentration is approximately equal to the solubility limit of hydrogen in niobium. As the hydrogen concentration exceeds C cr , ΔK th increases slowly as more hydrogen is added to the specimen. At load ratio 0.75, ΔK th decreases continuously as the hydrogen concentration is increased. The results provide evidence that two mechanisms are responsible for fatigue crack growth behavior in niobium-hydrogen alloys. First, embrittlement is retarded by hydride transformation-induced and plasticity-induced crack closures. Second, embrittlement is enhanced by the presence of hydrogen and hydride.