Faulted loops in neutron-irradiated zirconium

Faulted loops have been observed in high-purity zirconium irradiated at 723 K to 1.3 × 10²⁵ neutrons/m² (> 0.1 MeV). The transmission electron microscopy characterization of these $\text{1}\text{6}$ 〈202̄3〉 faulted loops on (0001) is described in detail. It was found that the faulted loops were invariably vacancy in character although the coexisting population of perfect $\text{1}\text{3}$ 〈112̄0〉 loops was of a mixed interstitial/vacancy nature. The faulted loops were observed in specimens of only two out of five batches of high-purity zirconium irradiated in this experiment. Even in these two specimens, the presence of faulted loops was restricted to the 723 K irradiation temperature; at 673 K only perfect $\text{1}\text{3}$ 〈112̄0〉 loops were seen.     

The characteristics of 15 MeV and fission neutron damage in niobium

Displacement damage by 15 MeV (d-Be source) and fission neutrons at 30°C in high purity niobium single crystals has been studied by transmission electron microscopy. The fluence of the 15 MeV neutrons was 1.8 × 10¹⁷n/cm² and the fluence of the fission neutrons (5 × 10¹⁷ n/cm²) was chosen so that samples from both types of irradiations had approximately the same damage energy. In both 15 MeV and fission neutron irradiated specimens, the loops were observed to be about $\text{2}\text{3}$ interstitial and $\text{1}\text{3}$ vacancy type. The analysis of Burgers vectors of the dislocation loops showed that more than $\text{2}\text{3}$ of the loops were perfect $\text{a}\text{2}\text{〈111〉}$ and that the rest were $\text{a}\text{2}\text{〈110〉}$ faulted. It is concluded that at equal damage energies, the detailed nature of the damage is the same for 15 MeV and fission neutron irradiated niobium.     

Chemical thermodynamic representations of 〈PuO2−x〉 and 〈U1−zPuzOw〉

All available oxygen potential-temperature-composition data for the calcium fluorite-structure 〈PuO_2?x〉^★★ phase were retrieved from the literature and utilized in the development of a binary solid solution representation of the phase. The data and phase relations are found to be best described by a solution of [Pu_{$\text{4}\text{3}$}O₂] and [PuO₂] with a temperature dependent interaction energy. The fluorite-structure 〈U_1?zPu_zO_w〉 is assumed to be represented by a combination of the binaries 〈PuO_2?x〉 and 〈UO₂ ± x〉, and thus treated as a solution of [$\text{Pu}{}_{\mathrm{<mtext>4</mtext><mtext>3</mtext>}}\text{O}{}_2$], [PuO₂], [UO₂], and either [U₂O_4.5] or [U₃O₇]. The resulting equations well reproduce the large amount of oxygen potential-temperature-composition data for the mixed oxide system, all of which were also retrieved from the literature. These models are the first that appear to display the appropriate oxygen potential-temperature-composition and phase relation behavior over the entire range of existence for the phases.     

The deposition of uranium-sesquicarbide

Deposition of U₂C₃ from an aic-melted mixture of UC + UC₂ ($\text{7.0 ± 0.1}$ wt % carbon) was examined primarily by the change of electrical resistivity as a function of time at fixed temperatures: 1370, 1440, 1575 and 1650°C. The rate constant of the reaction UC + UC₂ → U₂C₃ was investigated in detail. Growth with one constant dimension, which has been named layer growth, was predominate at 1370°C and during the early stages of growth at higher temperatures. Various growth schemes, which are generally involved, yield various activation energies. The rateconstant of thereaction was obtained as $\text{k}{}_{\mathrm{n}}\text{= 1.7 × 10}{}^{\mathrm{?2}}\text{exp(?n × 24.67/RT)}$ sec^?1, where $\text{n}$ depends on growth scheme and is usually a number less than 3. The activation energy of 24.67 kcal/mol in the above equation was obtained from rate constants with similar $\text{n}$ values ($\text{n ≈ 3}$) at 1440 and 1575°C.     

An interpretation of corduroy contrast in neutron irradiated zirconium

This paper describes an interpretation of a contrast effect that appears in electron micrographs of irradiated zirconium and Zircaloy. This ‘corduroy’ contrast has been analysed and shown to be an artefact associated with local relaxation of the thin foil in areas where there is pronounced alignment of the $\text{a/3 〈11}\text{2}\text{?}\text{0〉}$ loops.     

Void formation and solute segragation in Ti-14.4 at% Al

An alloy of Ti-14.4 at% Al was irradiated with 3.0 MeV ⁵⁸Ni ions at temperatures near 600°C and to damage levels from 1.3 to 14.7 dpa. TEM examination of these specimens showed the damage produced voids, dislocation loops and network dislocations. The undersize aluminum solute atom segregated to the voids, grain boundaries and the irradiated surface, causing enhanced precipitation of γ'-Ti₃Al at these sinks for point defects. Shells of $\text{α}{}_2\text{~ 75}\text{A}\text{?}$ thick coated the voids and a continuous film of γ' was observed on the irradiated surface and on some grain boundaries. The surface film of α₂ contained $\text{1}\text{3}\text{a 〈11}\text{2}\text{?}\text{0〉}$ antiphase boundaries. Voids have not been previously observed in irradiated hcp Ti and this is the first observation of substantial irradiation-induced segregation in an hcp alloy.     

A microstructural investigation of steady state creep of zircaloy-4 at 973 K

The creep behaviour of Zircaloy-4 at 973 K displays a transition at an applied stress of σ_T ≈ 25 MPa. In particular, only at stresses above this level does the loading strain exceed one elastic deflection, the primary creep strain vary systematically with stress or the strain rate at high strain approach a power law dependence. At stresses ? 25 MPa, steady state is not achieved even at strains of ~ 0.5.The present paper describes observations of the microstructures produced by creep. It is found that the average dislocation density p can be described by the empirical equation. σ = σ₀ + 0.9 Gb√ρ, where σ₀ = 10MPa. The dislocation is dominated by $\text{1}\text{3}\text{a〈11}\text{2}\text{?}\text{0〉}$ defects, which are most easily mobile on prismatic {011?0} planes, though evidence of the operation of basal slip and of the climb of $\text{1}\text{3}\text{a〈11}\text{2}\text{?}\text{0〉}$dislocations out of their prismatic planes is also presented.At all stresses above σ_T, well-developed subgrains are observed whose average size is inversely proportional to stress. At σ_T, the subgrain size is equal to the grain size of as-received material (~9 μm), though grain growth occurs during creep testing.A reduced power-law for the steady-state creep rate above σ_T is proposed ? ∝ (σ — σ_T)³, or, in terms of dislocation density $\text{? ∝ (ρ-ρ}{}_0\text{)}{}_{\mathrm{<mtext>3</mtext><mtext>2</mtext>}}$. The value of ρ₀is ~ 8 × 10^?12m^?2, which is suggested to represent the minimum dislocation density which is able to achieve pure dislocation strain, and which corresponds to the density at σ_T. At stresses below σ_T, the dislocation population is insufficient to allow the general grain shape change required by von Mises criterion. Only the more favourable slip systems on average are operative at all stresses above σ₀.Observations of the microstructure at applied stresses above σ_T are compatible with steady state creep being recovery controlled in that regime. At lower stress, some further deformation mechanism is required to act in conjunction with dislocation glide in order to achieve the observed strain while maintaining material continuity at grain boundaries.Observations of grain boundary structure suggest that this further mechanism involves long range grain boundary diffusion. The latter allows the movement of grain boundary dislocations, which in general will also cause migration and grain growth. Further, this is consistent firstly with the absence of steady state at low stresses, since the diffusion creep rate decreases as the grain size increases, and secondly with the equality of the subgrain and initial grain sizes at σ_T, since all the matrix dislocation accommodation processes must then involve grain boundary diffusion.     

Electrical resistivity of β-zirconium-oxygen solid solutions as a function of oxygen concentration and temperature

The electrical resistivity of recrystallized, pure β-zirconium at temperatures between 900 and 1700°C as well as the resistivity of β-zirconium-oxygen solid solutions at temperatures between 1100 and 1700°C have been measured. It has been found that the resistivity increases linearly with temperature at a constant oxygen concentration and increases linearly with the oxygen concentration at a constant temperature. The supplementary resistivity $\text{Δρ/Δ?}$ caused by the dissolution of oxygen decreases linearly with increasing temperature, i.e. Matthiessen's rule is not obeyed. The temperature (T) and oxygen concentration (?) dependence of the electrical resistivity ρ of β-zirconium-oxygen solid solutions at temperatures between 1100 and 1700°C can be represented by the relation ρ = 91.9 + 2.30 × 10^?2T + 10²f(3.75 ? 1.03 × 10^?3T) with ρ in μΩ · cm, ? in atomic ratio n_O/n_Zr and T in °C. Taking these results and literature data into consideration, a survey of the electrical resistivity of zirconium-oxygen solid solutions, including the α-range, is given.     

Autoirradiation d'un alliage cubique centre uranium-plutonium-molybdene

Measurements have been made of the length changes due to self-irradiation damage at 4.2 K for over 1300 h in a U-Pu-Mo bcc alloy (containing 20 at% plutonium and 15 at% molybdenum) stabilized by quenching. These are the first measurements of this type carried out on a bcc alloy. The initial rate of increase of δl/l was $\text{2,3 × 10}{}^{\mathrm{?7}}\text{h}{}^{\mathrm{?1}}$ If it is assumed that each α desintegration forms 1800 Frenkel pairs, it is found that Frenkel pairs have a formation volume equal to 0.5 atomic volume.     

Molecular dynamics simulation of low energy displacement cascades in Cu

Molecular dynamics computer simulations have been performed to study properties of low-energy (< 500 eV) displacement cascades in Cu. Various aspects of the time development of cascades are considered including instantaneous number of Frenkel pairs, partitioning of kinetic and potential energies, distribution of atom kinetic energies, cascade expansion rate, and Frenkel pair distributions. The anisotropy of the threshold energy for Frenkel-pair production is interpreted in terms of “branching”. Replacement sequences and the damage function are discussed based on analysis of events corresponding to 18 recoil directions. The damage function exhibits a plateau at $\text{v ~ 0.5}$ Frenkel pairs extending from 25–150 eV; at higher recoil energies the onset of multiple defect production is much slower than predicted by the modified Kinchin-Pease model.     

〈c〉-Component dislocations in zirconium alloys

Dislocations with $\text{〈c〉-}\text{component}$ Burgers vectors have been found in abundance near deformation twins, and to a lesser extent near grain boundary junctions, in deformed Zircaloy-2 and Zircaloy-4. Both pure $\text{〈c〉}$ and $\text{〈c + a〉}$ dislocations have been identified by TEM contrast experiments. The segments of $\text{〈c〉-}\text{component}$ dislocations tend to be long, straight, and to lie on either basal or pyramidal planes. It is suggested that these dislocations are generated in order to maintain compatibility between crystallites which differ significantly in their ability to accommodate an imposed deformation by $\text{〈a〉-}\text{slip}$. The manner in which $\text{〈c〉-}\text{component}$ dislocations can alter the partitioning of irradiation-produced point defects, and their influence on irradiation growth are discussed.     

On the unfaulting of vacancy loops in BCC metals

A review of radiation damage experiments on the bcc metals shows that differences in the crystallography of dislocation loops exist between the metals, and these differences can be important for the evolution of damage structure. This paper considers vacancy loops, for which the preferred Burgers vectors are observed to be $\text{1}\text{2}\text{〈111〉}$ in molybdenum and 〈100〉 in iron. All loops probably form from a common $\text{1}\text{2}\text{〈110〉}$ origin. The methods of computer simulation of discrete atomic crystallites and anisotropic elasticity theory are used to examine the energy and structure changes for the two unfaulting shears that produce the observed Burgers vectors. In computer simulation, unfaulting to 〈100〉 is favoured in both metals, and the elasticity calculations give the same result for loops which are only partially unfaulted. The apparent inconsistency between theory and experiment is discussed.     

Neutron irradiation effects in uranium dicarbides

Changes of electrical resistivity and lattice parameter in UC_1.96 after neutron irradiation from $\text{9 × 10}{}^{14}$ to $\text{2 × 10}{}^{18}$ nvt were studied. The resistivity was increased with the dose up to $\text{1 × 10}{}^7$ nvt, and saturated at that dose. Above 10¹⁸ nvt a steep increase was observed. In the lattice-parameter changes, on the other hand, a gradual increase was observed in the dose range between $\text{2 × 10}{}^{16}$ and $\text{8 × 10}{}^{17}$ nvt; above that dose, an abrupt increase followed. Annealing experiments on the resistivity were performed up to 1000°C using the specimens irradiated to the low dose of $\text{5 × 10}{}^{16}$ nvt, and the increased resistivity was completely recovered in three steps. The activation energies of each step were estimated to be 0.3, 0.5 and $\text{1.6 ± 0.2}$ eV.     

Infrared spectroscopic analysis of OH− and OD− in crystalline Li2O as a function of chemical treatment

Infrared spectroscopy has been used to study the chemical form and approximate concentration of OH^? and OD^? in Li₂O single crystals as a function of chemical treatment. Infrared absorption maxima at $\text{(3671±0.5)}\text{cm}{}^{\mathrm{?1}}\text{and}\text{(2711±3.3)}\text{cm}{}^{\mathrm{?1}}$ were observed for OH^? and OD^?, respectively. The absorption coefficient for OD^? was determined to be $\text{4.0±0.4}$ absorbance units per mol part per million OD^? per mm of sample thickness. Vacuum baking of Li₂O crystals reduced the OH^? and OD^? concentrations to <50 mppm; baking in a low moisture-level D₂ environment at 600 to 800°C appeared to lead to volatilization of LiD from the Li₂O crystals; and baking in D₂ containing ($\text{350±50}$) mppm D₂O at 600 to 800°C produced a measurable quantity of LiOD. In all cases, the observed spectra indicated the presence of only one distinguishable form of OH^? or OD^? in the Li₂O lattice. Because of the close correspondence of the observed absorption maxima to reported values for pure LiOH and LiOD, the most consistent (although not conclusive) interpretation is that the OH^? and OD^? are present as a separate LiOH or LiOD phase at room temperature. Only limited conclusions can be drawn regarding the chemical state of OH^? and OD^? during the elevated temperatures exposures. An estimate of the approximate value for the solubility of tritium in Li₂O at 800°C was made using data from D₂/Li₂O isothermal exposure experiments — this value was ? 25 wppm.     

Fission-enhanced self-diffusion of uranium in UO2 and UC

In-pile self-diffusion measurements in stoichiometric UO₂ sinters and single crystals and in arc-cast stoichiometric UC have been performed using the thin layer condition and ²³³U as tracer. The nominal irradiation temperature was 900°C. The resulting diffusion coefficients $\text{D1}$ of $\text{1.5 × 10}{}^{\mathrm{?16}}\text{cm}{}^2\text{· sec}{}^{\mathrm{?1}}$ for UO₂ and $\text{2.2 × 10}{}^{\mathrm{?17}}\text{cm}{}^2\text{· sec}{}^{\mathrm{?1}}$ for UC for a fission rate $\text{S}$ of 1 × 10¹³f/cm³ · sec represent radiation enhanced diffusion and are higher by factors of 10³ to 10⁴ than (extrapolated) coefficients of thermal diffusion. The data are of immediate relevance for understanding and predicting such important quantities as in-pile sintering and densification, diffusion controlled creep and fission gas behavior in the outer zones of the fuel. They are at the upper limit of expected values.     

The tensile deformation behavior of fine-grained α-plutonium

Tensile deformation of extruded monoclinic α-plutonium with an average grain size of 4 μm was studied at stress from 2 500 to 100 000 psi (17.3 to 689 MN/m²) and temperatures from 22 to 108°C. The strain rate varied from 10^?9 to 7 × 10^?3 sec^?1. The relation, $\text{? = 2.86 × 10}{}^{\mathrm{?7}}\text{σ}{}^{4.2}\text{exp}\text{(?25 600/}\text{RT}\text{)}\text{sec}{}^{\mathrm{?1}}$, was obeyed from 12 000 to 60 000 psi (71.7 to 414 MN/m²) for strain rates greater than about 10^?6 sec^?1. Stress and temperature dependences of creep rate over this stress range were in accord with a dislocation climb controlled creep model, although the power law behavior occurred at stresses higher than theory predicts. The value of 25 600 cal/mole proved a reasonable value for the activation energy for self-diffusion in α-plutonium. At lower stresses the apparent activation energy for creep increased with decreasing stress, and the stress exponent $\text{n}\text{(=}\text{d log}\text{?/}\text{d log}\text{σ)}$ increased from 4.2 to 7.9. The high apparent activation energies for creep and high n values at low stresses were attributed to grain growth during creep. Tensile elongation increased with decreasing strain rate and increasing temperature over the entire stress range. Low elongation at high stresses was attributed to lack of grain boundary sliding. Grain size changed during creep toward a size determined by stress. At the highest test temperatures and lowest stresses grain growth occurred during large strains, while at high stresses the average grain size decreased.     

Cavity nucleation calculations for irradiated metals

A theory of helium-assisted cavity nucleation in irradiated metals is modified and applied to conditions of continuous helium generation. The theory considers the nucleation and growth of cavities by coprecipitation of vacancies, interstitials, and inert gas atoms. Calculations are performed for type 304 stainless steel for comparison with ion irradiation experiments at $\text{~ 2 × 10}{}^{\mathrm{?4}}\text{dpa/s}$, with helium implantation at the rate of ~10^?2 appm/s, to a total damage of ~ 5 dpa, over the temperatures 773–973 K. Total cavity number density calculated ranges from 10²³ m^?3 at 773 K to 10²⁰ m^?3 at 973 K. The calculated incubation time for cavity appearance is 1000–3000 s (0.2–0.6 dpa). The calculated plot of cavity density versus time approximately reproduces the experimental data. Predicted cavity size distributions are roughly bell-shaped, but skewed in favor of larger cavity sizes. Calculated and experimental mean sizes agree within a factor of 3. The predictions of the model are found to change very little when most parameters are varied within reasonable limits. The model is, however, found to be strongly sensitive to cavity : matrix surface energy, as well as the rate that helium atoms are displaced from dislocations.     

Heterodiffusion de l'or,de l'argent et du cuivre dans les phases cubiques δ et ε du plutonium

The solute diffusion at infinite dilution of ¹⁹⁸Au and ^110mAg in cubic phases of Pu has been studied using the serial sectroning method. The solute diffusion coefficients in the b.c.c. ? phase can be expressed by: $\text{D}{}^{\mathrm{Au}}{}_{\mathrm{?Pu}}\text{= 5,7 × 10}{}^{\mathrm{?5}}$ exp(?10300/RT) cm²/s and $\text{D}{}^{\mathrm{Ag}}{}_{\mathrm{?Pu}}\text{= 4,9 × 10}{}^{\mathrm{?5}}\text{exp(?9600/RT) cm}{}^2\text{/s}$. The solute diffusion mechanism is interstitial of the dissociative type in both cases. These experiments confirm the activated interstitial model which has been proposed for self diffusion of ?Pu. Indeed the solute diffusion coefficients of Au and Ag are near of the self diffusion coefficients of Pu. The mechanisms are therefore interstitial in both cases. In the f.c.c. δ phase of Pu where self diffusion takes place by a vacancy mechanism, the solute diffusion coefficients of Au and Ag are near of the self diffusion coefficients of δ Pu. Solute diffusion takes place also by a vacancy mechanism. On the other hand, the extrapolation at infinite dilution of experiments of solute diffusion of Cu in ?Pu (Matano-Wagner coupling) gives the following results: $\text{D}{}^{\mathrm{Cu}}{}_{\mathrm{?Pu}}\text{= 1 × 10}{}^{\mathrm{?3}}\text{exp(?12300/RT)}$ cm²/s. The solute diffusion mechanism is interstitial of the dissociative type. In the ? phase the smaller the atomic radius the faster the migration: r_Co < r_Cu < r_?Pu < r_Ag = r_Au, and D^Co_?Pu > D^Cu_?Pu >D^Pu_?PU > D^Ag_?Pu ≈ D^Au_?Pu.     

Creep properties of oxide and carbide fuels under irradiation

The creep behaviour of uranium dioxide and uranium carbide has been examined in both bend and compression experiments in DIDO Materials Test Reactor. In UO₂ no significant variation in creep rate with dose and temperature occured above ~10²⁵ fissions m^?3 between 450°C and 1230°C, the high strain rates measured in compression at low doses being largely attributable to pore sintering. Both a linear rating and stress dependence were observed up to 40 MNm^?2 and creep rates were found to be independent of grain size. At higher doses ($\text{>6 × 10}{}^{26}\text{fissions m}{}^{\mathrm{?3}}$) transient strains were incurred on varying stress and temperature due to the development of grain boundary gas bubbles. This also resulted in a six fold increase in the radiation creep constant between $\text{6 × 10}{}^{26}$ and $\text{1.2 × 10}{}^{27}$ fissions m^?3. A similar pattern of behaviour with respect to rating and stress was observed in hyperstoichiometric UC between 450 and 800°C up to $\text{1 × 10}{}^{27}$ fissions m^?3. However the nominally steady state creep rate was a factor 8 lower than in UO₂ irradiated under the same conditions. The experimental results also suggest that the primary creep contribution to the initial strain in compression is much higher than in UO₂. There was no evidence of either transient strain on changing stress or of an increasing creep rate at high doses. The experimental observations are reported and discussed in relation to models for irradiation induced low temperature creep in ceramic fuels.     

Measurements of the polarization of radiation emitted by fast argon ions scattered by adsorbates on surfaces

The circular polarization of radiation emitted by fast Ar ions in the 4p'²F_{$\text{7}\text{2}$} state was measured following near-grazing collisions with a surface. Under uhv conditions, at 60 keV incident ion energy, the relative Stokes parameter $\text{S}\text{I}$ = 0.38, but with N₂ and H₂ gases present and adsorbed on the surface, $\text{S}\text{I}$ = 0.25 and 0.30, respectively. The original polarization was recovered following removal of the gases. These gas-specific, reversible polarization changes are interpreted in terms of formation and orientation of the fast ion state by a final collision involving electron capture from the adsorbates.