Aging effects on microstructural and mechanical properties of select refractory metal alloys for space-reactor applications

Refractory alloys based on niobium, tantalum and molybdenum are potential candidate materials for structural applications in proposed space nuclear reactors. Long-term microstructural stability is a requirement of these materials for their use in this type of creep dominated application. Early work on refractory metal alloys has shown aging embrittlement occurring for some niobium and tantalum-base alloys at temperatures near 40% of their melting temperatures in either the base metal or in weldments. Other work has suggested microstructural instabilities during long-term creep testing, leading to decreased creep performance. This paper examines the effect of aging 1100 h at 1098, 1248 and 1398 K on the microstructural and mechanical properties of two niobium (Nb-1Zr and FS-85), tantalum (T-111 and ASTAR-811C) and molybdenum (Mo-41Re and Mo-47.5Re) base alloys. Changes in material properties are examined through mechanical tensile testing coupled with electrical resistivity changes and microstructural examination through optical and electron microscopy analysis.     

Ga NMR and magnetic susceptibility in δ-phase of Pu1−xGax (x = 0.05, x = 0.08) alloys

⁶⁹Ga nuclear magnetic resonance spectra, line shifts (⁶⁹K) and nuclear spin-lattice relaxation rate have been measured in the 20 years aged Pu_0.95Ga_0.05 and in fresh prepared Pu_0.92Ga_0.08 alloys, stabilized δ-phase, at magnetic field of 9.4 T in the temperature range (10-500) K. The line shift and are determined correspondingly by the static and fluctuating-in-time parts of the local magnetic field that originates in transferred hyperfine coupling the Ga nuclear spin with the nearest f-electron environment of more magnetic Pu.Temperature behavior of the resonance properties is found the same in fresh Pu_0.92Ga_0.08 and aged Pu_0.95Ga_0.05 alloy. The NMR results are in favor that δ-phase of Pu_1−xGa_x alloys represents at T > 200 K the Kondo lattice, in which the localized electronic spins fluctuate independently from each other without any macroscopic coherence. The coherent state like in heavy-fermion liquids emerges in Pu_0.95Ga_0.05 below T^∗ = 200 K. A little bit higher estimate of crossover temperature T^∗ = 250 K was founded for Pu_0.92Ga_0.08.     

Behaviour of chromium steels in liquid Pb-55.5Bi with changing oxygen content and temperature

The behaviour of protective oxide layers on P122 steel and its welds and of ODS steel in liquid Pb_44.5Bi_55.5 (LBE) is examined under conditions of changing temperatures and oxygen concentrations. P122 (12Cr) and its welded joints are exposed to LBE at 550 °C for 4000 h with oxygen concentrations of 10⁻⁶ and 10⁻⁸ wt% (p(O₂) = 8.1 × 10⁻²³ bar and 5.2 × 10⁻²⁷ bar) which change every 800 h. It is found that like in case of constant oxygen concentration of 10⁻⁶ wt% a protective spinel layer (Fe(Fe_1−xCr_x)₂O₄) was maintained on P122 and also on its welded joint. Two experiments with exposure times of 4800 h are conducted on ODS steel, both with temperatures changing from 550 to 650 °C and back every 800 h, one experiment with 10⁻⁶ the other with 10⁻⁸ wt% oxygen in LBE. Both experiments show strong local dissolution attack after 4800 h which is in agreement with the behaviour of ODS in LBE at a constant temperature of 650 °C. However, dissolution attack is less in LBE with 10⁻⁸ wt% oxygen (p(O₂) = 3.0 × 10⁻²⁵ bar).     

Hydrogen and helium trapping in tungsten under single and sequential irradiations

The trapping of hydrogen and helium in polycrystalline tungsten irradiated with 500 eV He⁺, H⁺ and D⁺ ions, individually or sequentially, has been measured by thermal desorption spectroscopy. Specimens irradiated with 500 eV He⁺ at 300 K show three He release peaks in the vicinity of ∼500, ∼1000, and ∼1200 K. The helium is thought to form He vacancy complexes or bubbles. Increasing the specimen temperature to 700 K does not significantly affect the trapping behavior of He. Sequential He⁺-D⁺ irradiation at 300 K results in the elimination of He release above 800 K. Instead, both D and He were released in the range 400-800 K. This is interpreted as interstitial D and He released from the near surface. Sequential He⁺-D⁺ irradiation at 700 K resulted in a reduced single He peak at ∼1000 K with very little release observed below 800 K; no D was trapped for irradiations at 700 K. Sequential D⁺-He⁺ irradiations at 300 K show that He trapping occurs in much the same manner as for the He⁺-only case while D retention is reduced at the near surface. Sequential D⁺-He⁺ irradiations at 700 K indicate that pre-irradiation with D⁺ has little or no effect on the subsequent trapping behavior of He.     

Thermodynamic properties of ThxU1−xO2 (0 < x < 1) based on quantum-mechanical calculations and Monte-Carlo simulations

Th_xU_1−xO_2+y binary compositions occur in nature, uranothorianite, and as a mixed oxide nuclear fuel. As a nuclear fuel, important properties, such as the melting point, thermal conductivity, and the thermal expansion coefficient change as a function of composition. Additionally, for direct disposal of Th_xU_1−xO₂, the chemical durability changes as a function of composition, with the dissolution rate decreasing with increasing thoria content. UO₂ and ThO₂ have the same isometric structure, and the ionic radii of 8-fold coordinated U⁴⁺ and Th⁴⁺ are similar (1.14 nm and 1.19 nm, respectively). Thus, this binary is expected to form a complete solid solution. However, atomic-scale measurements or simulations of cation ordering and the associated thermodynamic properties of the Th_xU_1−xO₂ system have yet to be determined. A combination of density-functional theory, Monte-Carlo methods, and thermodynamic integration are used to calculate thermodynamic properties of the Th_xU_1−xO₂ binary (ΔH_mix, ΔG_mix, ΔS_mix, phase diagram). The Gibbs free energy of mixing (ΔG_mix) shows a miscibility gap at equilibration temperatures below 1000 K (e.g., E_exsoln = 0.13 kJ/(mol cations) at 750 K). Such a miscibility gap may indicate possible exsolution (i.e., phase separation upon cooling). A unique approach to evaluate the likelihood and kinetics of forming interfaces between U-rich and Th-rich has been chosen that compares the energy gain of forming separate phases with estimated energy losses of forming necessary interfaces. The result of such an approach is that the thermodynamic gain of phase separation does not overcome the increase in interface energy between exsolution lamellae for thin exsolution lamellae (10 Å). Lamella formation becomes energetically favorable with a reduction of the interface area and, thus, an increase in lamella thickness to >45 Å. However, this increase in lamellae thickness may be diffusion limited. Monte-Carlo simulations converge to an exsolved structure [lamellae || ] only for very low equilibration temperatures (below room temperature). In addition to the weak tendency to exsolve, there is an ordered arrangement of Th and U in the solid solution [alternating U and Th layers || {1 0 0}] that is energetically favored for the homogeneously mixed 50% Th configurations. Still, this tendency to order is so weak that ordering is seldom reached due to kinetic hindrances. The configurational entropy of mixing (ΔS_mix) is approximately equal to the point entropy at all temperatures, indicating that the system is not ordered.     

Charge transfer and delocalization studies from Kβ-to-Kα intensity ratios in CoxCu1−x alloys

Kβ-to-Kα X-ray intensity ratios of Co and Cu have been measured in pure metals and in alloys of Co_xCu_1−x (x = 0.8, 0.7, 0.6, 0.5, 0.4, 0.3 and 0.2) following excitation by 22.69 keV X-rays from a ¹⁰⁹Cd radioactive point source. The valence-electronic configurations of these metals were determined by corporation of measured Kβ-to-Kα X-ray intensity ratios with the results of multiconfiguration Dirac–Fock (MCDF) calculation for various valence-electronic configurations. Valance electronic configurations of Co and Cu in alloys indicate significant differences with respect to the pure metals. Our analysis indicates that these differences arise from delocalization and/or charge transfer phenomena in alloys. Namely the observed change of the valence-electronic configurations of metals in alloys can be explained with the transfer of 3d electrons from one element to the other element and/or the rearrangement of electrons between 3d and 4s, 4p states of individual metal atoms.     

Oxygen potential and thermal conductivity of (U, Pu) mixed oxides

(U, Pu) mixed oxides, (U_1−yPu_y)O_2−x, with y = 0.21 and 0.28 are being considered as fuels for the Prototype Fast Breeder Reactor (PFBR) in India. The use of urania-plutonia solid solutions in PFBR calls for accurate measurement of physicochemical properties of these materials. Hence, in the present study, oxygen potentials of (U_1−yPu_y)O_2−x, with y = 0.21 and 0.28 were measured over the temperature range 1073-1473 K covering an oxygen potential range of −550 to −300 kJ mol⁻¹ (O/M ratio from 1.96 to 2.000) by employing a H₂/H₂O gas equilibration technique followed by solid electrolyte EMFmeasurement. (U_1−yPu_y)O_2−x, with y = 0.40 is being used in the Fast Breeder Test Reactor (FBTR) in India to test the behaviour of fuels with high plutonium content. However, data on the oxygen potential as well as thermal conductivity of the mixed oxides with high plutonium content are scanty. Hence, the thermal diffusivity of (U_1−yPu_y)O₂, with y = 0.21, 0.28 and 0.40 was measured and the results of the measurements are reported.     

High-temperature steam-oxidation behavior of Zr-1Nb cladding alloy E110

Oxidation experiments were conducted at 1000-1200 °C in flowing steam with samples of as-received Zr-1Nb alloy E110 tubing and/or polished E110 tubing. The purpose was to determine the oxidation behavior of this alloy under postulated loss-of-coolant accident conditions in light water reactors. The as-received E110 tubing exhibited a high degree of susceptibility to nodular oxidation and breakaway oxidation at relatively low test times, as compared to other cladding alloys. The nodules grew much more rapidly at 1000 °C than 1100 °C, as did the associated hydrogen uptake. The oxidation behavior was strongly affected by the surface condition of the materials. Polishing to ≈0.1 μm roughness (the roughness of the as-received tubing was ≈0.4 μm) delayed breakaway oxidation. Polishing also removed surface impurities. For polished samples oxidized at 1100 °C, no significant nodular oxidation appeared up to 1000 s. For polished samples oxidized at 1000 °C, hydrogen uptake >100 wppm was delayed from ≈300 s to >900 s. Weight-gain coefficients were determined for pre-breakaway oxidation of polished-only and machined-and-polished E110 tubing samples: 0.162 (mg/cm²)/s^0.5 at 1000 °C and 0.613 (mg/cm²)/s^0.5 at 1100 °C.     

Effect of temperature on studtite stability: Thermogravimetry and differential scanning calorimetry investigations

The main objective of this work is the study of the influence of temperature on the stability of the uranyl peroxide tetrahydrate (UO₂O₂ · 4H₂O) studtite, which may form on the spent nuclear fuel surface as a secondary solid phase. Preliminary results on the synthesis of studtite in the laboratory at different temperatures have shown that the solid phases formed when mixing hydrogen peroxide and uranyl nitrate depends on temperature. Studtite is obtained at 298 K, meta-studtite (UO₂O₂ · 2H₂O) at 373 K, and meta-schoepite (UO₃ · nH₂O, with n < 2) at 423 K. Because of the temperature effect on the stability of uranyl peroxides, a thermogravimetric (TG) study of studtite has been performed. The main results obtained are that three transformations occur depending on temperature. At 403 K, studtite transforms to meta-studtite, at 504 K, meta-studtite transforms to meta-schoepite, and, finally, at 840 K, meta-schoepite transforms to U₃O₈. By means of the differential scanning calorimetry the molar enthalpies of the transformations occurring at 403 and 504 K have been determined to be −42 ± 10 and −46 ± 2 kJ mol⁻¹, respectively.     

Retention and surface blistering of helium irradiated tungsten as a first wall material

The first wall of an inertial fusion energy reactor may suffer from surface blistering and exfoliation due to helium ion irradiation and extreme temperatures. Tungsten is a candidate for the first wall material. A study of helium retention and surface blistering with regard to helium dose, temperature, pulsed implantation, and tungsten microstructure was conducted to better understand what may occur at the first wall of the reactor. Single crystal and polycrystalline tungsten samples were implanted with 1.3 MeV ³He in doses ranging from 10¹⁹ m⁻² to 10²² m⁻². Implanted samples were analyzed by ³He(d,p)⁴He nuclear reaction analysis and ³He(n,p)T neutron depth profiling techniques. Surface blistering was observed for doses greater than 10²¹ He/m². For He fluences of 5 × 10²⁰ He/m², similar retention levels in both microstructures resulted without blistering. Implantation and flash heating in cycles indicated that helium retention was mitigated with decreasing He dose per cycle.     

Surface morphology and deuterium retention in tungsten oxide layers exposed to low-energy, high flux D plasma

Surface morphology and deuterium retention in tungsten oxide layers (WO_3−z, z ? 0.25) grown on polycrystalline and recrystallized W substrates have been examined after exposure to a low-energy (38 eV/D), high flux (10²² D/m² s) D plasma to an ion fluence of 10²⁶ D/m² at various temperatures (up to ∼700 K). Characterization methods used were scanning electron microscopy, X-ray diffraction, Rutherford backscattering spectroscopy, and the D(³He,p)⁴He nuclear reaction analysis. During exposure to the D plasma at temperatures of 340-615 K, a partial reduction of the tungsten oxide takes place in the near-surface layer up to 0.3 μm in depth. Even at around room temperature, deuterium atoms diffuse several micrometers into the tungsten oxide. The high D concentration of about 0.1 D/W observed in the first micrometers below the surface at temperatures below 500 K can be related mainly to D atoms chemically bonded to O atoms. As the exposure temperature increases, the D concentration decreases, reaching about 2 × 10⁻⁴ D/W at 615 K. At plasma exposure temperatures of about 700 K, the oxide layer shrinks and loses a large fraction of oxygen.     

Thermal and mechanical properties of (U,Er)O2

Erbium is considered as a slow burnable poison suitable for use in light water reactors (LWRs). Addition of a small amount of Er₂O₃ to all UO₂ pellets will make it possible to develop super high burnup fuels in Japanese nuclear facilities which are now under the restriction of the upper limit of ²³⁵U enrichment. When utilizing the (U,Er)O₂ fuels, it is very important to understand the thermal and mechanical properties. Here we show the characterization results of (U_1−xEr_x)O₂ (0 ? x ? 0.1). We measured their thermal and mechanical properties and investigated the effect of Er addition on these properties of (U,Er)O₂. All Er completely dissolved in UO₂, and the lattice parameter decreased linearly with the Er content. Both the thermal conductivity and Young’s modulus of (U,Er)O₂ decreased with the Er content. These results would be useful for us in evaluating the performance of the (U,Er)O₂ fuels in LWRs.     

Light ion irradiation effects on stuffed Lu2(Ti2−xLux)O7−x/2 (x = 0, 0.4 and 0.67) structures

We have recently synthesized “stuffed” (i.e., excess Lu) Lu₂(Ti_2−xLu_x)O_7−x/2 (x = 0, 0.4 and 0.67) compounds using conventional ceramic processing. X-ray diffraction measurements indicate that stuffing more Lu³⁺ cations into the oxide structure leads eventually to an order-to-disorder (O-D) transition, from an ordered pyrochlore to a disordered fluorite crystal structure. At the maximum deviation in stoichiometry (x = 0.67), the Lu³⁺ and Ti⁴⁺ ions become completely randomized on the cation sublattices, and the oxygen “vacancies” are randomized on the anion sublattice. Samples were irradiated with 400 keV Ne²⁺ ions to fluences ranging from 1 × 10¹⁵ to 1 × 10¹⁶ ions/cm² at cryogenic temperatures (∼77 K). Ion irradiation effects in these samples were examined by using grazing incident X-ray diffraction. The results show that the ion irradiation tolerance increases with disordering extent in the non-stoichiometric Lu₂(Ti_2−xLu_x)O_7−x/2.     

The use of HI-ERDA/RBS and NRA/RBS to depth profile N in GaAs1−xNx thin films

N profiles of several GaAs_1−xN_x epitaxial layers with different N mole fractions in the range 0 < x < 0.14 were obtained by using (1) heavy-ion elastic recoil detection analysis (HI-ERDA) along with Rutherford backscattering spectrometry (RBS) using a 35 MeV Si⁶⁺ beam, and (2) nuclear reaction analysis (NRA) with the ¹⁴N(α, p)¹⁷O reaction, also with RBS, using a 3.7 MeV ⁴He⁺ beam. The results from the two techniques are compared and the advantages, disadvantages and capabilities are discussed.     

Enthalpy measurements of La2Te3O9 and La2Te4O11

Enthalpy increment measurements on La₂Te₃O₉(s) and La₂Te₄O₁₁(s) were carried out using a Calvet micro-calorimeter. The enthalpy values were analyzed using the non-linear curve fitting method. The dependence of enthalpy increments with temperature was given as: H°(T) − H°(298.15 K) (J mol⁻¹) = 360.70T + 0.00409T² + 133.568 × 10⁵/T − 149 923 (373 ? T (K) ? 936) for La₂Te₃O₉ and H°(T) − H°(298.15 K) (J mol⁻¹) = 331.927T + 0.0549T² + 29.3623 × 10⁵/T − 114 587 (373 ? T (K) ? 936) for La₂Te₄O₁₁.     

Kinetic and thermodynamic studies of the dissolution of thoria-urania solid solutions

The dissolution of Th_1−xU_xO₂ was investigated through leaching experiments combined with X-ray photoelectron spectroscopy (XPS) and X-ray absorption spectroscopy (XAS) analyses. These experiments were performed in acidic and in oxidizing conditions (nitric solutions), for several compositions of solid solutions ranging from x = 0.24 to 0.81. Static sequential experiments in acidic media performed at room temperature confirmed that higher concentration of uranium in the solid solution leads to higher release of uranium in the leachate whatever the pH. The normalized dissolution rate in oxidizing media is increasing all the more the content of uranium is increases in the mixed oxide. While for Th enriched solids, kinetic parameters remain similar to that of ThO₂, in the case of uranium enriched solids, a drastic change is observed, and kinetic parameters are similar to that of UO₂ ones. For x > 0.50, the saturation is reached in the leachate after 100 days. XPS and EXAFS analysis on leached samples pointed out an oxidation of U(IV) at the surface for x < 0.5, and in the bulk for x > 0.5. Enrichment in Th is also observed at the surface of the solid, indicating the formation of a protective layer of hydrated thorium oxide, or hydroxide. Finally, the solubility product of secondary phase was determined. The values obtained are in good agreement with that of ThO₂, Th(OH)₄ and ThO₂, xH₂O reported in the literature.     

Thermal transport properties of hafnium hydrides and deuterides

The thermal conductivities of δ′-, δ-, δ+ε-, and ε-phase hafnium hydrides and deuterides with various hydrogen isotope concentrations (HfH_x, 1.48 ? x ? 2.03; HfD_x, 1.55 ? x ? 1.94) were evaluated within the temperature range of 290-570 K from the measured thermal diffusivity, calculated specific heat, and density. The thermal conductivities of δ′-, δ-, δ+ε-, and ε-phase HfH_x and HfD_x are independent of the temperature within the range 300-550 K and are in the range 0.15-0.22 W/cm K and 0.17-0.23 W/cm K, respectively; these values are similar to and lower than the observed thermal conductivities of α-phase Hf. The experimental results for the electrical resistivities of δ′-, δ-, δ+ε-, and ε-phase HfH_x and HfD_x and the Lorenz number corresponding to the electronic conduction, obtained from the Wiedemann-Franz rule, indicated that heat conduction due to electron migration significantly influences the thermal conductivity values at high temperatures. On the other hand, heat conduction due to phonon migration significantly affects the isotope effects on the thermal transport properties.     

Electrochemical behaviour of gadolinium ion in molten LiCl-KCl eutectic

This work presents the electrochemical study of GdCl₃ in the molten LiCl-KCl eutectic in the temperature range 723-823 K. Transient electrochemical techniques such as cyclic voltammetry and chronopotentiometry, on an inert metallic tungsten working electrode, have been used in order to investigate the reduction mechanism and transport parameters. This study shows that Gd³⁺ ions are reduced to Gd metal by a single step mechanism with exchange of three electrons. Diffusion coefficient of GdCl₃ ions was determined at various temperatures, at 723 K the value is D = 0.88 10⁻⁵ cm² s⁻¹. Apparent standard reduction potential of the redox couple Gd³⁺/Gd has been determined by the open-circuit chronopotentiometry technique at several temperatures. Also the Gibbs free energy of GdCl₃ formation was determined and compared with thermodynamic data for pure compounds in the supercooled state in order to estimate the activity coefficient of Gd³⁺ in the molten LiCl-KCl eutectic.     

Testing IFE materials on Z

On a single-pulse basis, the tungsten armor for the chamber walls in a laser inertial fusion energy power plant must withstand X-ray fluences of 0.4-1.2 J/cm² with almost no mass loss, and preferably no surface changes. We have exposed preheated tungsten samples to 0.27 and 0.9 J/cm² X-ray fluence from the Z accelerator at Sandia National Laboratories to determine the single-shot X-ray damage threshold. Earlier focused ion beam analysis has shown that rolled powdered metal formed tungsten and tungsten alloys, will melt when exposed to 2.3 J/cm² on Z, but not at 1.3 J/cm². Three forms of tungsten - single-crystal (SING), chemical-vapor-deposited (CVD), and rolled powdered metal (PWM) - were exposed to fluence levels of 0.9 J/cm² without any apparent melting. However, the CVD and PWM sample surfaces were rougher after exposure than the SING sample, which was not roughened. BUCKY (1D) calculations show a threshold of 0.5 J/cm² for melting on Z. The present experiments indicate no melting but limited surface changes occur with polycrystalline samples (PWM and CVD) at 0.9 J/cm² and no surface changes other than debris for samples at 0.27 J/cm².     

Phase relations and linear thermal expansion of cubic solid solutions in the Th1−xMxO2−x/2 (M = Eu, Gd, Dy) systems

Cell parameters and linear thermal expansion studies of the Th-M oxide systems with general compositions Th_1−xM_xO_2−x/2 (M = Eu³⁺, Gd³⁺ and Dy³⁺, 0.0 ? x ? 1.0) are reported. The XRD patterns of each product were refined to specify the solid solubility limits of MO_1.5 in the ThO₂ lattice. The upper solid solubility limits of EuO_1.5, GdO_1.5 and DyO_1.5 in the ThO₂ lattice under conditions of slow cooling from 1673 K are represented as Th_0.50Eu_0.50O_1.75, Th_0.60Gd_0.40O_1.80 and Th_0.85Dy_0.15O_1.925, respectively. The linear thermal expansion (293-1123 K) of MO_1.5 and their single-phase solid solutions with thoria were investigated by dilatometery. The average linear thermal expansion coefficients () of the compounds decrease on going from EuO_1.5 to DyO_1.5. The values of for EuO_1.5, GdO_1.5 and DyO_1.5 containing solid solutions showed a downward trend as a function of the dopant concentration. The linear thermal expansion (293-1473 K) of the solid solutions investigated by high-temperature XRD also showed a similar trend.