Microstructural and mechanical property changes with aging of Mo-41Re and Mo-47.5Re alloys

The changes in microstructure and mechanical properties of Mo-41Re and Mo-47.5Re alloys were investigated following 1100 h thermal aging at 1098, 1248 and 1398 K. The electrical resistivity, hardness and tensile properties of the alloys were measured both before and after aging, along with the alloy microstructures though investigation by optical and electron microscopy techniques. The Mo-41Re alloy retained a single-phase solid solution microstructure following 1100 h aging at all temperatures, exhibiting no signs of precipitation, despite measurable changes in resistivity and hardness in the 1098 K aged material. Annealing Mo-47.5Re for 1 h at 1773 K resulted in a two-phase αMo + σ structure, with subsequent aging at 1398 K producing a further precipitation of the σ phase along the grain boundaries. This resulted in increases in resistivity, hardness and tensile strength with a corresponding reduction in ductility. Aging Mo-47.5Re at 1098 and 1248 K led to the development of the χ phase along grain boundaries, resulting in decreased resistivity and increased hardness and tensile strength while showing no loss in ductility relative to the as-annealed material.     

Microstructural and mechanical property changes in the Ta-base T-111 alloy following thermal aging

The microstructural changes occurring in the Ta-base T-111 (Ta-8W-2Hf) alloy during 1100 h thermal aging at 1098, 1248 and 1398 K under inert atmosphere and the influence on mechanical properties are reported. Electrical resistivity, hardness and tensile properties are compared between the as-annealed and aged conditions. Microstructural evaluations were performed by optical, scanning electron microscopy and transmission electron microscopy. An increase in the amount of grain boundary precipitation with increasing aging temperature was found to decrease the electrical resistivity and material strength. Precipitation at the grain boundaries was found to be a mixture of monoclinic and cubic structures, suggesting the development of mixed Hf oxides, carbides and nitrides. Precipitate development caused pronounced embrittlement of the alloy following aging at 1398 K.     

Radiation-damage in molybdenum-rhenium alloys for space reactor applications

Various Mo-Re alloys are attractive candidates for use as fuel cladding and core structural materials in spacecraft reactor applications. Molybdenum alloys with rhenium contents of 41-47.5% (wt%), in particular, have good creep resistance and ductility in both base metal and weldments. However, irradiation-induced changes such as transmutation and radiation-induced segregation could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to evaluate the performance of Mo-41Re and Mo-47.5Re after irradiation at space reactor relevant temperatures. Tensile specimens of Mo-41Re and Mo-47.5Re alloys were irradiated to ∼0.7 displacements per atom (dpa) at 1073, 1223, and 1373 K and ∼1.4 dpa at 1073 K in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Following irradiation, the specimens were strained to failure at a rate of 1 × 10⁻³ s⁻¹ in vacuum at the irradiation temperature. In addition, unirradiated specimens and specimens aged for 1100 h at each irradiation temperature were also tested. Fracture mode of the tensile specimens was determined. The tensile tests and fractography showed severe embrittlement and IG failure with increasing temperatures above 1100 K, even at the lowest fluence. This high temperature embrittlement is likely the result of irradiation-induced changes such as transmutation and radiation-induced segregation. These factors could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to examine the irradiation-induced degradation for these Mo-Re alloys under neutron irradiation.     

Optimization of resistance spot welding on the assembly of refractory alloy 50Mo-50Re thin sheet

Resistance spot welding (RSW) was employed to pre-join refractory alloy 50Mo-50Re (wt%) sheet with a 0.127 mm gage. Five important welding parameters (hold time, electrode, ramp time, weld current and electrode force) were adjusted in an attempt to optimize the welding quality. It was found that increasing the hold time from 50 ms to 999 ms improved the weld strength. Use of rod-shaped electrodes produced symmetric nugget and enhanced the weld strength. Use of a ramp time of 8 ms minimized electrode sticking and molten metal expulsion. The weld strength continuously increased with increasing the weld current up to 1100 A, but the probabilities of occurrence of electrode sticking and molten metal expulsion were also increased. Electrode force was increased from 4.44 N to 17.8 N, in order to reduce the inconsistency of the welding quality. Welding defects including porosities, columnar grains and composition segregation were also studied.     

Effect of sintering conditions on the microstructure and mechanical properties of ZrN as a surrogate for actinide nitride fuels

Pellets of sintered ZrN were studied to optimize the mechanical properties and microstructures needed in nitride fuel pellets, using ZrN as a surrogate for actinide nitrides and as potential component in low fertile and inert matrix fuels. Samples were prepared via sintering in either Ar or N₂ (with and without 6% H₂) and at 1300 °C or 1600 °C. A significant difference in the hardness was measured ranging from 1000 (Kg/mm²) in samples sintered at 1600 °C in argon to 100 (Kg/mm²) in samples sintered at 1300 °C in nitrogen. Samples with 6% hydrogen added to the sintering environment experienced a decrease in hardness, as well as an increase in intergranular cracking as compared to samples sintered without hydrogen, suggesting hydrogen embrittlement. Grain size was more uniform in samples sintered in pure Ar as compared to Ar-H₂, while the latter had a larger fraction of high angle grain boundaries than the former. Cracking around indents had a clear tendency to follow high angle boundaries, which were found to be intrinsically weak in ZrN.     

Literature review of thermal and radiation performance parameters for high-temperature, uranium dioxide fueled cermet materials

High-temperature fissile-fueled cermet literature was reviewed. Data are presented primarily for the W-UO₂ as this was the system most frequently studied; other reviewed systems include cermets with Mo, Re, or alloys as a matrix. Failure mechanisms for the cermets are typically degradation of mechanical integrity and loss of fuel. Mechanical failure can occur through stresses produced from dissimilar expansion coefficients, voids created from diffusion of dissimilar materials or formation of metal hydride and subsequent volume expansion. Fuel loss failure can occur by high temperature surface vaporization or by vaporization after loss of mechanical integrity. Techniques found to aid in retaining fuel include the use of coatings around UO₂ fuel particles, use of oxide stabilizers in the UO₂, minimizing grain sizes in the metal matrix, minimizing impurities, controlling the cermet sintering atmosphere, and cladding around the cermet.     

Alloying and microstructure stability in the high-temperature Mo-Si-B system

Multi-phase alloys in the Mo-Si-B system are identified as high-temperature structural materials due to their high melting points (above 2000 °C) and excellent oxidation resistance attributed to the self-healing characteristics of borosilica layer up to 1400 °C. In the current study, the effect of alloying additions to achieve a reduced weight density has been examined in terms of changes in the microstructure and phase stability. The critical factor underlying the microstructural changes is related to the influence of the alloying additions on the stability of the high melting temperature ternary-based Mo₅SiB₂ (T₂) borosilicide phase.     

Influence of δ phase precipitation on the stress corrosion cracking resistance of alloy 718 in PWR primary water

The negative influence of δ phase on the intergranular stress corrosion cracking (IGSCC) resistance of alloy 718 is commonly taken for granted. In addition, δ phase formed at low temperature (about 1023 K) do not present the same characteristics than the one formed at higher temperatures (from 1173 to 1273 K). The aim of the present study is then to understand how δ phase precipitation could enhance crack initiation in alloy 718, whatever the form of δ phase is. For that purpose, several heat treatments leading to δ phase precipitation were realized on two alloy 718 heats, one sensitive to IGSCC and the second not. Specific slow strain rate tensile tests carried out on thin tensile specimens in simulated PWR primary medium at 633 K conclusively prove that δ phase has no effect on the intrinsic sensitivity to intergranular crack initiation of tested heats.     

Electron beam irradiation effects on the mechanical, thermal and surface properties of a fluoroelastomer

Fluoroelastomer can be used as a sealing material for different purposes. The aim of this work is the evaluation of the effects of the ionizing radiation of an electron beam (EB) on the mechanical, thermal and surface properties of a commercial fluoroelastomer containing carbon black and inorganic fillers. The material was irradiated with overall doses between 10 and 250 kGy. Tensile strength (stress and strain at break), hardness (Shore A) and compression set were evaluated. Thermal behavior was evaluated by thermogravimetric analysis and differential scanning calorimetry. Surface modifications were inspected using scanning electron microscopy (SEM) and optical microscopy. The experiments have shown that EB irradiation promotes beneficial changes in the fluoroelastomer tensile strength behavior while compression set remain constant and the glass transition temperature increases. The SEM micrographs have shown compactness in the irradiated samples, although optical observations showed no surface morphology changes.     

Influence of Portevin-Le Chatelier effect on rupture mode of alloy 718 specimens

A set of tensile tests has been carried out under air environment in the temperature range (450-700 °C) in order to characterize the interactions between oxidation assisted intergranular cracking and Portevin-Le Chatelier effect. It is shown that the occurrence of jerky flow stops the intergranular damaging mechanism. The construction of a temperature versus strain rate diagram showing dynamic strain aging, intergranular fracture and PLC instabilities locations is then proposed. The surprising relation between crack initiation disappearance and PLC instabilities of type C in the temperature range explored is discussed. Finally the assumption of the occurrence of such phenomenon at the tip of a propagating crack is addressed.     

Neutron irradiation of sapphire for compressive strengthening. II. Physical properties changes

Irradiation of sapphire with fast neutrons (0.8-10 MeV) at a fluence of 10²²/m² increased the c-axis compressive strength and the c-plane biaxial flexure strength at 600 °C by a factor of ∼2.5. Both effects are attributed to inhibition of r-plane twin propagation by damage clusters resulting from neutron impact. The a-plane biaxial flexure strength and four-point flexure strength in the c- and m-directions decreased by 10-23% at 600 °C after neutron irradiation. Neutron irradiation had little or no effect on thermal conductivity, infrared absorption, elastic constants, hardness, and fracture toughness. A featureless electron paramagnetic resonance signal at g=2.02 was correlated with the strength increase: This signal grew in amplitude with increasing neutron irradiation, which also increased the compressive strength. Annealing conditions that reversed the strengthening also annihilated the g=2.02 signal. A signal associated with a paramagnetic center containing two Al nuclei was not correlated with strength. Ultraviolet and visible color centers also were not correlated with strength in that they could be removed by annealing at temperatures that were too low to reverse the compressive strengthening effect of neutron irradiation.     

Radiation cross-linking in ultra-high molecular weight polyethylene for orthopaedic applications

The motivation for radiation cross-linking of ultra-high molecular weight polyethylene (UHMWPE) is to increase its wear resistance to be used as bearing surfaces for total joint arthroplasty. However, radiation also leaves behind long-lived residual free radicals in this polymer, the reactions of which can detrimentally affect mechanical properties. In this review, we focus on the radiation cross-linking and oxidative stability of first and second generation highly cross-linked UHMWPEs developed in our laboratory.     

Development of oxides dispersion strengthened steels for high temperature nuclear reactor applications

By introducing a dispersion of nanosized yttrium oxides particles into a steel matrix, the upper temperature limit in mechanical creep strength can be enhanced in temperature by 100 K at least. Production routes for the production of a new class of oxides dispersion strengthened (ODS) steels are investigated within this work. Preliminary results obtained when doping pure iron matrix phase with two types of yttrium oxides (Y₂O₃) nanoparticles (commercial as well as laboratory fabricated nanopowder) are presented. The twofold purpose of this work is firstly to obtain a comparative analysis between the commercial and the laboratory fabricated Y₂O₃ nanopowder used to produce the doped iron, and secondly to demonstrate the feasibility of new production route by observing the nanostructure of the first test batches with pure iron. Observations are carried out with transmission electron microscopy (TEM) to determine the size distribution of the particles in the powder, while glow discharge optical emission spectroscopy (GDOES) and high resolution-scanning electron microscopy (HR-SEM) are used to analyze the chemical composition and the homogeneity of the produced doped iron. It is demonstrated, that even with small size particles nanopowder fabricated in the laboratory, the distribution is fairly homogeneous compared to the one obtained with a relatively large particles commercial nanopowder, confirming the feasibility of the new production route.     

Modeling and simulation of irradiation hardening in structural ferritic steels for advanced nuclear reactors

Hardening and embrittlement are controlled by interactions between dislocations and irradiation induced defect clusters. In this work we employ the visco plastic self consistent (VPSC) polycrystalline code in order to model the yield stress dependence in ferritic steels on the irradiation dose. We implement the dispersed barrier hardening model in the VPSC code by introducing a hardening law, function of the strain, to describe the threshold resolved shear stress required to activate dislocations. The size and number density of the defect clusters varies with the irradiation dose in the model. We find that VPSC calculations show excellent agreement with the experimental data set. Such modeling efforts can both reproduce experimental data and also guide future experiments of irradiation hardening.     

Heat load behaviors of plasma sprayed tungsten coatings on copper alloys with different compliant layers

Plasma sprayed tungsten (PS-W) coatings with the compliant layers of titanium (Ti), nickel-chromium-aluminum (NiCrAl) alloys and W/Cu mixtures were fabricated on copper alloys, and their properties of the porosity, oxygen content, thermal conductivity and bonding strength were measured. High heat flux tests of actively cooled W coatings were performed by means of an electron beam facility. The results indicated that APS-W coating showed a poorer heat transfer capability and thermo-mechanical properties than VPS-W coating, and the compliant layers improved W coating performance under the heat flux load. Among three compliant layers, W/Cu was the preferable because of its better effects on heat removal and stress alleviating. The optimization of W/Cu compliant layer found that 0.1 mm and 25 vol.%W was optimum compliant layer structure for 1 mm W coating, which induced a 23% reduction of the maximum stress compared to the sharp interface, and the plastic strain was reduced to 0.01% from 1.55%.     

Neutron irradiation of sapphire for compressive strengthening.: I. Processing conditions and compressive strength

Sapphire suffers a dramatic loss of c-axis compression strength at elevated temperatures. Irradiation of sapphire with fission-spectrum neutrons to an exposure of ∼10²² neutrons/m² in the core of a 1 MW fission reactor increased the c-axis compression strength by a factor of ∼3 at 600 °C. Strength was similarly improved when 99% of slow neutrons (?0.1 eV) were removed by 10B and Cd shields during irradiation. Annealing at 600 °C for 10 min changed the yellow-brown color of irradiated sapphire to pale yellow, but had no effect on compressive strength. Annealing irradiated sapphire at 1200 °C for 24 h reduced the compressive strength to its baseline value. Transmission electron microscopy suggests that fast-neutron-induced displacement damage inhibits the propagation of r-plane twins which are responsible for the low compressive strength. When irradiated with 10B and Cd shielding, sapphire that was not grown in iridium crucibles is safe for unrestricted handling after 1 month.     

Degradation of mechanical properties of structural reactor materials induced by formation of stress concentrators

The paper deals with the crack nucleation and stability in strain fields of stress concentrators (e.g. voids, gas bubbles, secondary phase precipitates). A general equation describes critical and subcritical crack length as a function of external (applied loading) and internal (stress concentrator type, normal traction, elastic properties of matrix, etc.) parameters. For the critical crack an analog of the Griffith criterion is found. The reduction of fracture stress due to different types of internal stress concentrators was evaluated.     

Effect of thermo-mechanical processing on microstructure and creep properties of the foils of alloy 617

The effect of rolling and annealing on the microstructure and high temperature creep properties of alloy 617 were investigated. Two types of foil specimens with different thickness reductions were prepared by thermo-mechanical processing. Recrystallization and grain growth were readily observed at specimens annealed at 950 and 1100 °C. The uniform coarse grains increase resistance against creep deformation. The grain size effect in creep deformation was dominant up to 900 °C, while dynamic recrystallization effect became dominant at 1000 °C. Dynamic recrystallization was observed in all the creep deformed foils, even though some specimens had already been (statically) recrystallized during annealing. Steady state creep rates decreased with increasing annealing temperature in the less rolled foils. The apparent activation energy Q_app for the creep deformation increased from 271 to 361 kJ/mol as the annealing temperature increased from 950 to 1100 °C.     

Mechanical properties of Eurofer 97 in Pb-16Li and irradiation effect

To be used in a fusion reactor, structural materials, and in particular steels, has to be selected and optimised in their composition to achieve a reduction in the long-term radioactive waste. A reduction in the long-term radioactive inventory could be reached substituting elements like molybdenum, niobium and nickel with other ones like tantalum and tungsten which have the same functions as alloying elements and, if irradiated, do not produce long lived radioisotopes. The martensitic steel belonging to the family of 8-9% Cr Eurofer 97 is considered the reference structural steel for fusion application. However, only few information are available about its mechanical properties in the liquid eutectic alloy Pb-16%Li. Particularly, the problem of liquid metal embrittlement (LME) has not been studied in detail and the effect of neutron irradiation on LME has not been investigated at all so far. This work presents the results obtained irradiating tensile specimens of Eurofer 97 up to 5.9 dpa in lead lithium. Tensile tests of samples have been performed out of pile in the same alloy at the same temperature at which irradiation was carried out.