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DOE OSTI2014

Defect interactions with stepped CeO{sub 2}/SrTiO{sub 3} interfaces: Implications for radiation damage evolution and fast ion conduction

Due to reduced dimensions and increased interfacial content, nanocomposite oxides offer improved functionalities in a wide variety of advanced technological applications, including their potential use as radiation tolerant materials. To better understand the role of interface structures in influencing the radiation damage tolerance of oxides, we have conducted atomistic calculations to elucidate the behavior of radiation-induced point defects (vacancies and interstitials) at interface steps in a model CeO{sub 2}/SrTiO{sub 3} system. We find that atomic-scale steps at the interface have substantial influence on the defect behavior, which ultimately dictate the material performance in hostile irradiation environments. Distinctive steps react dissimilarly to cation and anion defects, effectively becoming biased sinks for different types of defects. Steps also attract cation interstitials, leaving behind an excess of immobile vacancies. Further, defects introduce significant structural and chemical distortions primarily at the steps. These two factors are plausible origins for the enhanced amorphization at steps seen in our recent experiments. The present work indicates that comprehensive examination of the interaction of radiation-induced point defects with the atomic-scale topology and defect structure of heterointerfaces is essential to evaluate the radiation tolerance of nanocomposites. Finally, our results have implications for other applications, such as fast ion conduction.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2015

Impact of implanted phosphorus on the diffusivity of boron and its applicability to silicon solar cells

Boron diffusivity reduction in extrinsically doped silicon was investigated in the context of a process combination consisting of BBr{sub 3} furnace diffusion and preceding Phosphorus ion implantation. The implantation of Phosphorus leads to a substantial blocking of Boron during the subsequent Boron diffusion. First, the influences of ion implantation induced point defects as well as the initial P doping on B diffusivity were studied independently. Here, it was found that not the defects created during ion implantation but the P doping itself results in the observed B diffusion retardation. The influence of the initial P concentration was investigated in more detail by varying the P implantation dose. A secondary ion mass spectrometry (SIMS) analysis of the BSG layer after the B diffusion revealed that the B diffusion retardation is not due to potential P content in the BSG layer but rather caused by the n-type doping of the crystalline silicon itself. Based on the observations the B diffusion retardation was classified into three groups: (i) no reduction of B diffusivity, (ii) reduced B diffusivity, and (iii) blocking of the B diffusion. The retardation of B diffusion can well be explained by the phosphorus doping level resulting in a Fermi level shift and pairing of B and P ions, both reducing the B diffusivity. Besides these main influences, there are probably additional transient phenomena responsible for the blocking of boron. Those might be an interstitial transport mechanism caused by P diffusion that reduces interstitial concentration at the surface or the silicon/BSG interface shift due to oxidation during the BBr{sub 3} diffusion process. Lifetime measurements revealed that the residual (non-blocked) B leads to an increased dark saturation current density in the P doped region. Nevertheless, electrical quality is on a high level and was further increased by reducing the B dose as well as by removing the first few nanometers of the silicon surface after the BBr{sub 3} diffusion.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2015

Point defect states in Sb-doped germanium

Defect states in n-type Sb-doped germanium were investigated by deep-level transient spectroscopy. Cobalt-60 gamma rays were used to generate isolated vacancies and interstitials which diffuse and react with impurities in the material to form four defect states (E{sub 37}, E{sub 30}, E{sub 22}, and E{sub 21}) in the upper half of the bandgap. Irradiations at 77 K and 300 K as well as isothermal anneals were performed to characterize the relationships between the four observable defects. E{sub 37} is assigned to the Sb donor-vacancy associate (E-center) and is the only vacancy containing defect giving an estimate of 2 × 10{sup 11 }cm{sup −3} Mrad{sup −1} for the uncorrelated vacancy-interstitial pair introduction rate. The remaining three defect states are interstitial associates and transform among one another. Conversion ratios between E{sub 22}, E{sub 21}, and E{sub 30} indicate that E{sub 22} likely contains two interstitials.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2016

First principles calculations of point defect diffusion in CdS buffer layers: Implications for Cu(In,Ga)(Se,S){sub 2} and Cu{sub 2}ZnSn(Se,S){sub 4}-based thin-film photovoltaics

We investigate point defects in CdS buffer layers that may arise from intermixing with Cu(In,Ga)Se{sub 2} (CIGSe) or Cu{sub 2}ZnSn(S,Se){sub 4} (CZTSSe) absorber layers in thin-film photovoltaics (PV). Using hybrid functional calculations, we characterize the migration barriers of Cu, In, Ga, Se, Sn, Zn, Na, and K impurities and assess the activation energies necessary for their diffusion into the bulk of the buffer. We find that Cu, In, and Ga are the most mobile defects in CIGS-derived impurities, with diffusion expected to proceed into the buffer via interstitial-hopping and cadmium vacancy-assisted mechanisms at temperatures ∼400 °C. Cu is predicted to strongly favor migration paths within the basal plane of the wurtzite CdS lattice, which may facilitate defect clustering and ultimately the formation of Cu-rich interfacial phases as observed by energy dispersive x-ray spectroscopic elemental maps in real PV devices. Se, Zn, and Sn defects are found to exhibit much larger activation energies and are not expected to diffuse within the CdS bulk at temperatures compatible with typical PV processing temperatures. Lastly, we find that Na interstitials are expected to exhibit slightly lower activation energies than K interstitials despite having a larger migration barrier. Still, we find both alkali species are expected to diffuse via an interstitially mediated mechanism at slightly higher temperatures than enable In, Ga, and Cu diffusion in the bulk. Our results indicate that processing temperatures in excess of ∼400 °C will lead to more interfacial intermixing with CdS buffer layers in CIGSe devices, and less so for CZTSSe absorbers where only Cu is expected to significantly diffuse into the buffer.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Radiation-induced bistable centers with deep levels in silicon n{sup +}–p structures

The method of deep level transient spectroscopy is used to study electrically active defects in p-type silicon crystals irradiated with MeV electrons and α particles. A new radiation-induced defect with the properties of bistable centers is determined and studied. After keeping the irradiated samples at room temperature for a long time or after their short-time annealing at T ∼ 370 K, this defect does not display any electrical activity in p-type silicon. However, as a result of the subsequent injection of minority charge carriers, this center transforms into the metastable configuration with deep levels located at E{sub V} + 0.45 and E{sub V} + 0.54 eV. The reverse transition to the main configuration occurs in the temperature range of 50–100°C and is characterized by the activation energy ∼1.25 eV and a frequency factor of ∼5 × 10{sup 15} s{sup –1}. The determined defect is thermally stable at temperatures as high as T ∼ 450 K. It is assumed that this defect can either be a complex of an intrinsic interstitial silicon atom with an interstitial carbon atom or a complex consisting of an intrinsic interstitial silicon atom with an interstitial boron atom.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Generalized Rate Theory for Void and Bubble Swelling and its Application to Plutonium Metal Alloys

In the classical rate theory for void swelling, vacancies and self-interstitials are produced by radiation in equal numbers, and in addition, thermal vacancies are also generated at the sinks, primarily at edge dislocations, at voids, and at grain boundaries. In contrast, due to the high formation energy of self-interstitials for normal metals and alloys, their thermal generation is negligible, as pointed out by Bullough and Perrin. However, recent DFT calculations of the formation energy of self-interstitial atoms in bcc metals have revealed that the sum of formation and migration energies for self-interstitials atoms (SIA) is of the same order of magnitude as for vacancies. The ratio of the activation energies for thermal generation of SIA and vacancies is presented. For fcc metals, this ratio is around three, but for bcc metals it is around 1.5. Reviewing theoretical predictions of point defect properties in δ-Pu, this ratio could possibly be less than one. As a result, thermal generation of SIA in bcc metals and in plutonium must be taken into considerations when modeling the growth of voids and of helium bubbles, and the classical rate theory (CRT) for void and bubble swelling must be extended to a generalized rate theory (GRT).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Which parameters might help to predict an impact on reactivity for mixed fissile units in a storage array?

Several studies have already been performed on mixed fissile units in storage arrays and concluded that the neutron spectrum of the fissile material, the shape of fissile units, the interstitial moderation and structural materials may have an impact on reactivity. For this paper, calculations are performed on a simplified infinite planar array with one level of fissile material units axially reflected by concrete. Fissile material is modelled by a cylinder shape centered in the array mesh selected (120 cm height and 60 cm pitch). Either air or interstitial moderation by water is modelled between the fissile units. Three parameters are studied: the fissile materials (with various moderation conditions), the cylinder dimension and the nature of the structural material. The main following conclusions can be drawn from these calculations. First, for several mixed-array cases, air-spaced mixed arrays present a high increase on reactivity whereas interstitial moderation between fissile units leads to no or little impact. Secondly, fissile unit shapes do not explain directly the mixing impact on reactivity. Thirdly, structural materials have an important impact on reactivity in mixed array (either an increase or a decrease) due to their absorbing and/or moderating properties either when air or interstitial moderation are modelled between fissile units. Fourthly, no simple parameter was found to predict the impact on reactivity of mixed array in all studied configurations. However, a comparison of different neutron spectrum parameters can help to predict the absence or potential presence of a 'mixing effect'. Indeed, if all parameters are similar, no 'mixing effect' is observed. But, if these parameters present differences, there might be a 'mixing effect'. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Kinetic Monte Carlo Simulation of Hydrogen Diffusion in Tungsten

Nuclear fusion is considered as the most promising sustainable energy source. If achieved, it could provide electricity to the mankind for millions of years. During the reaction, the fusion ash is continuously removed from the magnetic gap, the divertor region, and the divertor facing component is subjected to intense plasma bombardment. It is proposed to use tungsten as the divertor facing material, because it has low physical sputtering yield and superior thermomechanical properties. However, the high mobility of hydrogen in tungsten and the high bombardment flux bring the hydrogen isotopes deep into the tungsten tile. The presence of hydrogen in tungsten tends to weaken the gram boundaries and produce surface blistering. The ejection of tungsten particulates into the plasma core will cool the reaction and cease the fusion reactions. Therefore, it is important to have a detailed understanding of the hydrogen diffusion in tungsten bulk. The KMC method is a stochastic computer simulation to predict the time evolution of a certain system. Typically, at a given time, the system will evolve through a few possible paths with known probabilities. In each step, the KMC method analyzes the entire system and randomly determines the system evolving events based on all transition rates of all possible events and advance the physical time. This process is repeated till the simulation of the entire physical time is completed. In previous work, the authors developed a kinetic Monte Carlo (KMC) algorithm and program to simulate hydrogen diffusion process on tungsten reconstructed surface. In this article, this KMC method is extended to model the hydrogen diffusion in tungsten bulk. The diffusion coefficients are calculated and fitted to the Arrhenius equation. The result matches the experiment measurement very well. KMC diffusion simulation needs the locations of the hydrogen trapping sites, diffusion paths and the corresponding diffusion energy barriers. The hydrogen interstitial sites in tungsten crystal have been thoroughly studied by density function theory (DFT) and molecular dynamics (MD). Johnson and Carter used DFT to determine the preferred location of the hydrogen interstitial sites m tungsten crystal, which are the tetrahedral sites (T-sites) in body-centered-cubic (bcc) lattice surfaces. All T-sites are located on the surfaces of a lattice cell. There are four T-sites per surface and twelve T-sites per lattice cell. The hydrogen atoms diffuse by traveling from one T-site to one of its nearest four T-site neighbors. Thus, each T-site is surrounded by two neighbors on the same surface, and one neighbor on each of the two adjacent surfaces perpendicular to the surface holding the center T-site. The diffusion barrier between two neighboring T-sites was calculated to be 0.38 eV. Knowing the diffusion path, energy barrier, and the crystal structure, the KMC simulation is readily to be performed. Based on other researchers' DFT calculations, the hydrogen stable interstitial site in tungsten bulk is determined as the tetrahedral site on the surface of the bee lattice. Each T-site has 4 neighbors and the energy barrier between two adjacent T-sites are estimated to be 0.38 eV. Based on those facts and the KMC theory, a MATLAB program is developed to model the hydrogen diffusion in tungsten bulk. In the MATLAB program, substrate temperature, hydrogen concentration, physical time to be simulated, and the number of cycles are all customizable. To compare the KMC determined diffusion coefficients with experiment results in the literature, the simulations are performed for temperature 300 K - 2500 K and low hydrogen concentration (0.1%), mimicking the experiment condition. The calculated diffusion coefficients match the reference values measured by the experiment under low hydrogen concentration condition. The results are also fitted to the Arrhenius equation. The pre-exponential factor and the average activation energy in Arrhenius equation are obtained. The good agreement between simulation and experiment proves that the developed KMC algorithm truly reflect the well-established theory of interstitial hydrogens in tungsten. The developed program could easily be used to model the diffusion process and predict the diffusion coefficients under other concentrations and temperatures. This work is part of the ongoing development of hydrogen diffusion in tungsten integrating surface diffusion (previous work), bulk diffusion (this work), and adsorption/desorption on surface (future work)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Scaling laws and stability of nano-sized defect clusters in niobium via atomistic simulations and statistical analysis

The predictions of scaling laws for the structure and properties of defect clusters are generally limited to small defect clusters in their ground-state configurations. We investigated the size and geometrical configuration dependence of nano-sized defect clusters in niobium (Nb) using molecular dynamics. We studied the structure and stability of large clusters of size up to fifty defects for vacancies and one hundred defects for interstitials, as well as the role of helium and metastable configurations on the stability of these clusters. We compared three different interatomic potentials in order to determine the relative stability of these clusters as a function of their size and geometrical configurations. Additionally, we conducted a statistical analysis to predict the formation and binding energies of interstitial clusters as a function of both their size and configuration. We find that the size dependence of vacancy and interstitial clusters can be approximated by functional forms that account for bulk and surface effects as well as some considerations of elastic interactions. We also find that helium and metastable configurations can make vacancy and interstitial clusters thermally stable depending on the configuration. Our parameterized functional forms for the formation and binding energies are valid for a very broad range of defect sizes and configurations making it possible to be used directly in a coarse-grained modeling strategy such as Monte Carlo, cluster dynamics or dislocation dynamics which look at defect accumulation and evolution in microstructures.

36 MATERIALS SCIENCE↗

DOE OSTI2014

Investigation of oxygen point defects in cubic ZrO 2 by density functional theory

The energetics of formation and migration of the oxygen vacancy and interstitial in cubic ZrO 2 are investigated by density functional theory calculations. In an O-rich environment, the negatively charged oxygen interstitial is the most dominant defect whereas, the positively charged oxygen vacancy is the most dominant defect under O-poor conditions. Oxygen interstitial migration occurs by the interstitialcy and the direct interstitial mechanisms, with calculated migration energy barriers of 2.94 eV and 2.15 eV, respectively. For the oxygen vacancy, diffusion is preferred along the <100> direction, and the calculated energy barriers are 0.26 eV for V 2+ O , 0.27 eV for V 1+ O and 0.54 eV for V 0 O . Furthermore, these results indicate that oxygen diffusivity is higher through the vacancy-migration mechanism.

36 MATERIALS SCIENCE↗

DOE OSTI2017

First-principles study of fission gas incorporation and migration in zirconium nitride

To evaluate the effectiveness of ZrN as a diffusion barrier against fission gases, we investigate in this paper the incorporation and migration of fission gas atoms, with a focus on Xe, in ZrN by first-principles calculations. The formations of point defects in ZrN, including vacancies, interstitials, divacancies, Frenkel pairs, and Schottky defects, are first studied. Among all the defects, the Schottky defect with two vacancies as first nearest neighbor is predicted to be the most favorable incorporation site for fission gas Xe in ZrN. The migration of Xe gas atom in ZrN is investigated through two diffusion mechanisms, i.e., interstitial and vacancy-assisted diffusions. The migration barrier of Xe gas atom through the intrinsic interstitials in ZrN is considerably lower than that through vacancies. Finally, therefore, at low temperatures fission gas Xe atoms diffuse mainly through interstitials in single crystal ZrN, whereas at high temperatures Xe may diffuse in ZrN assisted by vacancies.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Impact of amorphous pockets on displacement damage evolution in silicon

Silicon has long been known to exhibit amorphization in response to heavy particle bombardment. For doses below the total amorphization threshold, partial amorphization is observed in the form of scattered amorphous pockets. While extensive research has gone into modeling the formation and evolution of amorphous pockets in response to irradiation, no studies yet investigate their impact on the evolution of other damage such as interstitial supersaturation and clustering. In this study, we survey the impact of amorphous pockets on defect evolution in silicon when treated as static sinks. MD is first used to show that amorphous pockets provide energetically favorable sites for point defects relative to the crystalline bulk, supporting the hypothesis that they act as sinks. A 0-D cluster dynamics model is then constructed, taking an interstitial clustering model from the literature and including amorphous pockets as a sink species. We conduct our survey for temperatures between 30 and 400 °C and sink strengths between 1 to 6 x 10 10 cm −2 . Both implantation- and radiation-induced damage states are investigated using interstitial and vacancy concentrations as initial condition variables. We find that, due to the differing migration rates of the interstitial and the vacancy, amorphous pockets have a non-monotonic impact on the final damage state depending on the effective sink strength of the amorphous pockets, resulting in increased damage formation in regimes of intermediate amorphization. In conclusion, this result emphasizes the important role of amorphous pockets in governing the evolution of damage in partially amorphized crystalline materials.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Coupled cluster and dislocation dynamics modeling of microstructure evolution in irradiated materials

We develop here a coupled cluster and dislocation dynamics framework to study the microstructure evolution of irradiated materials. The framework not only accounts for the three dimensional diffusion of radiation-generated clusters, but also their interaction with dislocation networks and the resultant climb motion of discrete dislocations within finite crystals. The framework is solved with a superposition solution scheme, and is applied to investigate the evolution of the irradiation-induced dislocation loops in zirconium (Zr), considering the effects of various bias factors including the diffusion anisotropy difference (DAD) of interstitials and interstitial clusters, the dislocation bias of defects to discrete dislocation segments, and the production bias of defects from the radiation cascade. We find that the DAD is the most critical factor influencing the kinetics of the loop evolution in Zr, while the recombination/interaction of mobile defects can induce a strong spatial dependence of the loop evolution together with the DAD. Here, the method is also adopted to study the evolution of interstitial $\langle$a$\rangle$ and vacancy $\langle$c$\rangle$ dislocation loop ensembles consistent with the microstructure observed during irradiation-induced growth of Zr. Our findings not only reveal the spatial dependence of the size and ellipticity of the dislocation loops, but also suggest a limit on the anisotropy factor of interstitials to reproduce the co-growth of $\langle$a$\rangle$ and $\langle$c$\rangle$ loops in zirconium, in good agreement with experimental observations and other simulation results.

Bias factors↗

DOE OSTI2016

Roles of Vacancy/Interstitial Diffusion and Segregation in the Microchemistry at Grain Boundaries of Irradiated Fe-Cr-Ni alloys

A detailed analysis of the diffusion fluxes near and at grain boundaries of irradiated Fe–Cr–Ni alloys, induced by preferential atom-vacancy and atom-interstitial coupling, is presented. The diffusion flux equations were based on the Perks model formulated through the linear theory of the thermodynamics of irreversible processes. The preferential atom-vacancy coupling was described by the mobility model, whereas the preferential atom-interstitial coupling was described by the interstitial binding model. The composition dependence of the thermodynamic factor was modeled using the CALPHAD approach. The calculated fluxes up to 10 dpa suggested the dominant diffusion mechanism for chromium and iron is via vacancy, while that for nickel can swing from the vacancy to the interstitial dominant mechanism. The diffusion flux in the vicinity of a grain boundary was found to be greatly modified by the segregation induced by irradiation, leading to the oscillatory behavior of alloy compositions in this region.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

Defect cluster and nonequilibrium gas bubble associated growth in irradiated UMo fuels – a cluster dynamics and phase field model

Irradiation examination shows that gas bubble swelling kinetics is much faster after irradiation-induced recrystallization than that prior recrystallization in UMo fuels. It implies that gas bubbles in coarse grains and small recrystallized grains have different growth behavior. In this work, a phase-field model of gas bubble evolution integrating microstructure dependent cluster dynamics has been developed, for the first time, to study the gas bubble swelling behavior in the recrystallization zone of UMo fuels. Generation, diffusion, reaction, sink, emission and clustering of vacancies and interstitials are described by the cluster dynamics model while a phase-field model is used to describe the evolution of non-equilibrium gas bubbles including nucleation and growth. With the coupled model, the effect of defect generation rate, clustering rate, interstitial emission and sink rates on grain boundaries on the gas bubble evolution are systematically simulated. A set of model parameters (defect generation rate, clustering rate, interstitial emission and sink rates) is determined by comparing measured and simulated gas bubble swelling kinetics. The results demonstrate that interstitial clustering is one of the important physical mechanisms which results in a fast gas bubble swelling kinetics in the recrystallization zone. The developed model can also be extended to study the associated growth of defect and second phase precipitates often observed in irradiated materials.

Hu, Shenyang↗

DOE OSTI2024

Parameterization of vacancy production rate in phase-field models of fission gas bubble evolution in nuclear fuel

Phase-field modeling has increasingly been used to study microstructural evolution in fission gas bubbles in nuclear fuel to improve understanding of fission gas release. To improve computational efficiency, often only vacancies and gas atoms are included as defect species. In this case, the net effects of vacancy and interstitial production, recombination, and biased sink absorption are included as a net vacancy source, or net vacancy source combined with an effective sink. However, there has been a lack of clarity on what parameter values should be used for these approaches to best match the more complete physical picture that includes interstitials and vacancies. Here, we compare a phase-field model of void growth to analytical models for the source-only and source plus sink approach to gain insight into how the phase-field models can be parameterized effectively. The source-only approach provides greater flexibility to match growth rates determined from the full vacancy-interstitial picture. A strategy was developed for determining the value of the net vacancy source term by comparing to an analytical model that includes vacancy and interstitial production, recombination, and biased sink absorption. Finally, this strategy can be used to parameterize phase-field models of fission gas bubble growth.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Synthesis, structure and electrochemical properties of layered La{sub 2}Li{sub 2x}(CO{sub 3}){sub 1−x}O{sub 2+2x}

An oxide ion conductor, La{sub 2}Li{sub 2x}(CO{sub 3}){sub 1−x}O{sub 2+x} (x=0.15), was synthesized using a solid-state reaction, and its structure was subsequently determined by Rietveld refinement analyses of synchrotron X-ray and high-resolution powder neutron diffraction data. These analyses determined that carbonate groups are situated between (La{sub 2}O{sub 2}{sup 2+}){sub n} layers in an ordered arrangement and that both oxygen vacancies and interstitial oxygen ions exist within the carbonate layers, resulting in a composition of La{sub 2}Li{sub 0.3}(CO{sub 3}){sub 0.85}O{sub 2.3}. This material is structurally stable up to 500 °C and exhibits a conductivity of 5.5×10{sup −5} S cm{sup −1} at 450 °C. Oxide vacancies associated with carbonate anions appear to be responsible for the ionic conduction observed in this material. - Graphical abstract: Thermal ellipsoid projection of the carbonate layer viewed from the c-axis, showing oxygen migration paths: (A, blue line) interstitial hopping and (B, green line) co-operative rotation. Display Omitted - Highlights: • Lithium insertion in the structure causes the structural changes in carbonate layer. • These structural changes create oxygen vacancies within the flexible carbonate layers and introduce interstitial oxygen. • These interstitial oxygen atoms and oxygen vacancies are involved in oxide ion diffusion through the carbonate layer.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2024

Formation of large loop-hydrogen complexes and related effects on mechanical properties of Zirconium investigated with molecular dynamics method

Here, using molecular dynamics (MD) method, interaction between hydrogen (H) atoms and interstitial or vacancy dislocation loops in Zr is extensively studied. The results show that the binding of hydrogen atoms to dislocation loops is anisotropic (attractive or repulsive), depending on the position of the hydrogen atoms with respect to the loop’s core. The anisotropy is governed by excess volume and stress field induced by the loop. For a 1/3[$2\bar1$$\bar1$$0$] interstitial dislocation loop, hydrogen atoms preferentially segregate near the outside of the loop and close to the decomposition knots of the loop. While for a 1/3[$2\bar1$$\bar1$$0$] or 1/6[$20$$\bar2$$3$] vacancy loop, hydrogen atoms prefer to segregate near the inside of the loop. The present work indicates large loop-hydrogen complexes could form in Zr. The binding of hydrogen to loops is stable up to 600 K, while dissociation is observed at 900 K. Formation of loop-hydrogen complexes affect the activation of slip system, resulting in higher yield stress of Zr. The effect increases with the increasing ratio of H to vacancies or interstitials in the vacancy or interstitial loops. The results imply the effects of large loop-hydrogen complexes should be considered to fully understand radiation damage in Zr.

36 MATERIALS SCIENCE↗