H in α-Zr and in zirconium hydrides: solubility, effect on dimensional changes, and the role of defects (original) (raw)

Hydrogen storage in some ternary and quaternary zirconium-based alloys with the C14 structure

Journal of the Less Common Metals, 1985

The Zr~.s%.NnCrl.2s, ZrMnFeT, (T = Cr, Ni, Co; 0 < 3c < 0.4) and ZrCrFe, + X (0 < x < 0.5) alloys were studied for their hydrogen storage characteristics in the temperature range 23-150 "C and hydrogen equilibrium pressure (0.01-50 atm). The kinetics of hydrogen sorption by these alloys are observed to be extremely rapid, and the processes are almost complete in 60-200 s. Hydrogen capacities of these alloys are in the range 92-222 cm3 H, (g alloy)-I. The hyperstoichiometric elements appear to substitute at the zirconium site in these AB,-type alloys with a net effect of raising the decomposition pressure of their hydrides several-fold over that of the ZrMn, and ZrCr, hydrides. The ZrMnFeCo,.4 alloy does not absorb measurable quantities of hydrogen, presumably because of the high decomposition pressure of its hypothetical hydride. The effectiveness of alloying elements in destabilizing the ZrMn, hydride increases in the following order: Cr < Mn < Fe < Ni < Co. It is found that, unlike ZrMn,, ZrCr, cannot be made hyperstoichiometric with chromium or manganese. The hydrides of ZrCrFe 1 + .+ alloys have very favorable dissociation pressures, e.g. 0.9-2 atm at room temperature. The enthalpies and entropies of dehydrogenation of their hydrides are in the range 20-23 kJ (mol H,))' and 66-80 J mall' K-l respectively, which are advantageous for application purposes.

Hydrogen in zirconium alloys: A review

Journal of Nuclear Materials, 2019

Hydrogen absorbed into zirconium alloy nuclear fuel cladding as a result of the waterside corrosion reaction can affect the properties of nuclear fuel, principally through the precipitation of brittle hydride particles. Multiple phenomena are involved in this overall process: after hydrogen pickup degradation of mechanical properties is controlled by hydrogen transport, hydride precipitation and dissolution kinetics and the formation of specific mesoscale hydride microstructures. The precipitation of hydrides especially affects cladding ductility and fracture toughness, but can also affect other phenomena, including via stress-induced hydride reorientation. These processes can affect cladding performance both during normal operation and during extended dry storage, as hydride morphology can be modified during the preparatory vacuum drying processes. We review the processes of hydrogen transport, hydride precipitation and dissolution and formation of mesoscale hydride microstructures, and highlight where more research is needed, both from an experimental and from a modeling point of view.

Structure and Thermodynamical Properties of Zirconium Hydrids from First Principle

2011

Zirconium alloys are used as nuclear fuel cladding material due to their mechanical and corrosion resistant properties together with their favorable cross-section for neutron scattering. At running conditions, however, there will be an increase of hydrogen in the vicinity of the cladding surface at the water side of the fuel. The hydrogen will diffuse into the cladding material and at certain conditions, such as lower temperatures and external load, hydrides will precipitate out in the material and cause well known embrittlement, blistering and other unwanted effects. Using phase-field methods it is now possible to model precipitation buildup in metals, for example as a function of hydrogen concentration, temperature and external load, but the technique relies on input of parameters, such as the formation energy of the hydrides and matrix. To that end, we have computed, using the density functional theory (DFT) code GPAW, the latent heat of fusion as well as solved the crystal structure for three zirconium hydride polymorphs: δ-ZrH 1.6 , γ-ZrH, and -ZrH 2 .

Atomic-scale Ab-initio study of the Zr-H system: I. Bulk properties

Acta Materialia, 2002

Bulk properties of the Zr-H system were studied in the framework of the density functional theory. The local density approximation (LDA) is found to be insufficient for a proper description of interactions between Zr and H atoms and the generalized gradient approximation (GGA) is required. In αZr, H atoms preferentially occupy tetrahedral (T) sites at low temperatures, and can be regarded as being independent of each other up to very short distances, except for repulsive interactions between dumbbells in the same interstitial site. The Zr density of electronic states is perturbed by the presence of H, which induces the emergence of localized states. H diffusion occurs along the c → axis preferentially in octahedral (O) sites, and in the basal plane by alternate jumps into T and O sites. In the γ(ZrH), δ (ZrH 1.5) and ε(ZrH 2) hydrides, H-H interactions cannot be neglected, the nearly equal formation energies of these compounds indicate that their relative stabilities probably depend on mechanical and thermal contributions to free energies, and in fcc Zr, H atoms tend to adopt planar arrangements for compositions close to ZrH.

The role of chemical free energy and elastic strain in the nucleation of zirconium hydride

Journal of Nuclear Materials, 2013

In this work a combination of synchrotron X-ray diffraction and thermodynamic modelling has been used to study the dissolution and precipitation of zirconium hydride in a-Zr establishing the role of elastic misfit strain and chemical free energy in the a ? a + d phase transformation. The nucleation of zirconium hydride is dominated by the chemical free energy where the chemical driving force for hydride precipitation is proportional to the terminal-solid solubility for precipitation and can be predicted by a function that is analogous to the universal nucleation parameter for the bainite transformation in ferrous alloys. The terminal-solid solubility for precipitation was found to be kinetically limited P287°C at a cooling rate of 5°C min À1 or greater. The terminal solubilities were established using an offset method applied to the lattice strain data where a resolution of $10 wppm H can be achieved in the hci-direction. This is aided by the introduction of intra-granular strains in the hci-direction during cooling as a result of the thermal expansion anisotropy which increases the anisotropy associated with the misfitting H atoms within the a-Zr lattice.

Hydrides of ZrMn2-based alloys substoichiometric in zirconium for engineering applications

Journal of the Less Common Metals, 1985

The hydrogen storage characteristics of TiMn I .s and Zr, _ XT, Mn, Fe, (T = Ti, Mn, Fe) alloys substoichiometric in zirconium in the range 0 < x < 0.3, with manganese sites on ZrMn, lattice substituted by iron such that y + z = 1, were studied. Pressure-composition isotherms and other thermodynamic data are presented. All the host alloys and their hydrides exhibit a Cl4 (MgZn,) structure. The hydrogen capacity of the alloys is large and the hydrides have very favorable dissociation pressures. The work computed for the hydrogen charging and discharging of the Zr, _ xTxMnr Fe, alloys was observed to be lower than that for TiMni, and other accepted hydrogen storage materials. Utilization of the Zr, ~ xTXMn,,Fe, alloys in a hydrogenpowered automobile and in a heat pump or refrigerator appears to have distinct advantages.

Adsorption and diffusion of hydrogen in Zircaloy-4

Hydrogen in zirconium alloys is considered in many nuclear safety issues. Below 500°C, rather limited knowledge is available on the combined hydrogen adsorption at the sample surface and diffusion in the metal. A modeling of hydrogen gaseous charging has been established starting with a set of relevant laws and parameters derived from open literature. Simulating the hydrogen charging process requires simultaneous analysis of gaseous surface adsorption, hydrogen solid-solution diffusion and precipitation, when exceeding the material solubility limit. The modeling has been extended to reproduce the solid-gas exchange. Gaseous charging experiments have been performed at 420°C on Stress Relieved Annealed (SRA) Zircaloy-4 cladding samples to validate the model. The sample hydrogen content has been systematically measured after charging and compared to the calculated value thus providing a validation of the adsorption modeling. Complementary tests have been carried out on Recrystallized Annealed (RXA) Zircaloy-4 rods to characterize the combined diffusion and adsorption process. The hydrogen concentration distribution has been characterized using an inverse technique based on destructive analyses of the samples. This additional set of data was relevant for the validation of the hydrogen combined adsorption/diffusion modeling up to 420°C.

Molecular Dynamics Study of Hydrogen in α -Zirconium

International Journal of Nuclear Energy, 2014

Molecular dynamics approach is used to simulate hydrogen (H) diffusion in zirconium. Zirconium alloys are used in fuel channels of many nuclear reactors. Previously developed embedded atom method (EAM) and modified embedded atom method (MEAM) are tested and a good agreement with experimental data for lattice parameters, cohesive energy, and mechanical properties is obtained. Both EAM and MEAM are used to calculate hydrogen diffusion in zirconium. At higher temperatures and in the presence of hydrogen, MEAM calculation predicts an unstable zirconium structure and low diffusion coefficients. Mean square displacement (MSD) of hydrogen in bulk zirconium is calculated at a temperature range of 500-1200 K with diffusion coefficient at 500 K equals 1.92 * 10 −7 cm 2 /sec and at 1200 K has a value 1.47 * 10 −4 cm 2 /sec. Activation energy of hydrogen diffusion calculated using Arrhenius plot was found to be 11.3 kcal/mol which is in agreement with published experimental results. Hydrogen diffusion is the highest along basal planes of hexagonal close packed zirconium.