Search

Solution annealing and age hardening are important processes for achieving good engineering and chemical properties of alloy 718. The composition of alloy 718 also plays an important role as it can affect both mechanical (γ”- and γ’-phase formation) and irradiation behaviors. Therefore, five different sets of alloy 718 samples with two different chemical compositions were fabricated using different processing conditions and irradiated up to 4 dpa. Based on the effects of irradiation on the microstructural data of γ”- and γ’-phases, irradiation-induced dislocation statistics, and crystallographic data, it can be inferred that the samples processed at high solution annealing temperature (1093°C) behave better under irradiation compared to the samples processed at low solution annealing temperatures (945 and 1065°C).

This work presents 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. 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.

Alloy 718 is generally considered a highly corrosion-resistant material but can still be susceptible to stress corrosion cracking (SCC). The combination of factors leading to SCC susceptibility in the alloy is not always clear enough. In the present work, alloy 718 leaf spring (LS) materials that suffered stress corrosion damage during two 24-month cycles in pressurized water reactor service, operated to >45 MWd/mtU burn-up, was investigated. Compared to archival samples fabricated through the same processing conditions, little microstructural and property changes occurred in the material with in-service irradiation, contrary to high dose rate laboratory-based experiments reported in literature. Though the lack of delta phase formation along grain boundaries would suggest a more SCC resistant microstructure, grain boundary cracking in the material was extensive. Crack propagation routes were explored through focused ion beam milling of specimens near the crack tip for transmission electron microscopy as well as in polished plan view and cross-sectional samples for electron backscatter diffraction analysis. It has been shown in this study that cracks propagated mainly along random high-angle grain boundaries, with the material around cracks displaying a high local density of dislocations. The slip lines were produced through the local deformation of the leaf spring material above their yield strength. The cause for local SCC appears to be related to oxidation of both slip lines and grain boundaries, which under the high in-service stresses resulted in crack development in the material.

Analytical equations were developed to calculate the strength factors of precipitates, Frank loops, and cavities in austenitic alloys, which strongly depend on barrier type, size, geometry and density, as well as temperature. Calculated strength factors were successfully used to estimate radiation hardening using the broadly employed dispersed barrier-hardening model, leading to good agreement with experimentally measured hardening in neutron-irradiated type 304 and 316 stainless steel variants. The formulated strength factor provides a route for more reliable hardening predictions and can be easily incorporated into component simulations and design.

The dynamics of deformation localization and dislocation channel formation were investigated in situ in a neutron-irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction, and transmission electron microscopy (TEM). Channel formation was observed at ∼70% of the polycrystalline yield stress of the irradiated materials (σ0.2). It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the σ0.2, channels often formed near the middle of the grain boundary. For a single grain, the role of elastic stiffness value (Young’s modulus) in channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in “soft” grains with a high Schmid factor located near “stiff” grains with high elastic stiffness. The spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one-third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. In the AISI 304 steel, channels in grains oriented close to 〈0 0 1〉||TA (tensile axis) and 〈1 0 1〉||TA were twin free and grain with 〈1 1 1〉||TA and grains oriented close to a Schmid factor maximum contained deformation twins.

A specific phenomenon – highly localized regions of deformation – was found and investigated at the free surface and near-surface layer of a neutron irradiated AISI 304 stainless steel bend specimen deformed to a maximum surface strain of 0.8%. It was shown that local plastic deformation near the surface might reach significant levels being localized at specific spots even when the maximum free surface strain remains below 1%. The effect was not observed in non-irradiated steel of the same composition at similar strain levels. Cross-sectional EBSD analysis demonstrated that the local misorientation level was highest near the free surface and diminished with increasing depth in these regions. (S)TEM indicated that the local density of dislocation channels might vary up to an order of magnitude. These channels may contain twins or may be twin free depending on grain orientation and local strain levels. BCC-phase (α-martensite) formation associated with channel-grain boundary intersection points was observed using EBSD and STEM in the near-surface layer.

This work aims at utilizing modern computational microstructural modeling tools to accelerate the understanding of phase stability in austenitic steels under extended thermal aging. Using the CALPHAD approach, a thermodynamic database OCTANT (ORNL Computational Thermodynamics for Applied Nuclear Technology), including elements of Fe, C, Cr, Ni, Mn, Mo, Si, and Ti, has been developed with a focus on reliable thermodynamic modeling of precipitate phases in AISI 316 austenitic stainless steels. The thermodynamic database was validated by comparing the calculated results with experimental data from commercial 316 austenitic steels. The developed computational thermodynamics was then coupled with precipitation kinetics simulation to understand the temporal evolution of precipitates in austenitic steels under long-term thermal aging (up to 600,000 h) at a temperature regime from 300 to 900 °C. This study discusses the effect of dislocation density and difusion coefficients on the precipitation kinetics at low temperatures, which shed a light on investigating the phase stability and transformation in austenitic steels used in light water reactors.

Iron-based austenitic alloys are often observed to develop magnetic properties during irradiation, possibly associated with the radiation-induced acceleration of ferrite phase formation. Some of the parametric sensitivities of this phenomenon have been addressed using a series of alloys irradiated in the BOR-60 reactor at 593 K. An increase in magnetic phase amount for all alloys was observed over the 0–12 dpa dose range. However, magnetic phase (ferrite according to TEM results) did not appear to continuously increase at higher doses (above 12 dpa) but did tend to saturate. The formation of a magnetic phase in austenitic stainless steels during irradiation at 593 K appeared to be sensitive to alloy composition. It was found that silicon and manganese accelerated ferrite accumulation in the dose range of 0–12 dpa, whereas carbon and probably molybdenum resisted it. Also, an increase in grain size resisted ferrite formation, but cold work was found to stimulate it.

Surface relief due to localized deformation in a 4.4-dpa neutron-irradiated AISI 304 stainless steel was investigated using scanning electron microscopy coupled with electron backscattering diffraction and scanning transmission electron microscopy. It was found a body-centered-cubic (BCC) phase (deformation-induced martensite) had formed at the surface of the deformed specimen along the steps generated from dislocation channels. Martensitic hill-like formations with widths of ∼1 μm and depths of several microns were observed at channels with heights greater than ∼150 nm above the original surface. Martensite at dislocation channels was observed in grains along the [0 0 1]–[1 1 1] orientation but not in those along the [1 0 1] orientation.

RAFM steels are one of the candidate structural materials for fusion reactors, in which tantalum (Ta) and tungsten (W) are alloyed to replace niobium (Nb) and molybdenum (Mo) in conventional FM steels, respectively. This paper, using three RAFM heats, presents the effects of Ta and the primary austenite stabilizer carbon (C) on microstructure and strength. Thermomechanical treatment (TMT) was also applied to the heats, leading to significant increases in strength, attributable to the TMT-refined sub-grains and precipitates. The Ta-alloying favored the formation of (V, Ta)(N, C) and (Ta, V)C and exhibited greater strength. Fractographs also revealed the beneficial effects of TMT and Ta-alloying. However, extra C content, favoring a larger amount of M23C6 precipitates, did not show strengthening effect.

Life extension of the existing nuclear power plants imposes significant challenges to core structural materials that suffer increased fluences. This paper presents the microstructural evolution of a type 304 stainless steel and its variants alloyed with extra Ni and Cr under neutron irradiation at ∼320 °C for up to 10.2 dpa. Similar to the reported data of type 304 variants, a large amount of Frank loops, ultrafine G-phase/M23C6 particles, and limited amount of cavities were observed in the irradiated samples. The irradiation promoted the growth of pre-existing M23C6 at grain boundaries and resulted in some phase transformation to CrC in the alloy with both extra Ni and Cr. A new type of ultrafine precipitates, possibly (Ti,Cr)N, was observed in all the samples, and its amount was increased by the irradiation. Additionally, α-ferrite was observed in the type 304 steel but not in the Ni or Ni + Cr alloyed variants. The effect of Ni and Cr alloying on the microstructural evolution is discussed.

The objective of this work is to explore possible thermodynamic correlations between the degradation modes of austenitic and ferritic–martensitic alloys observed in high temperature He–CO–CO 2 environments with oxygen and carbon chemical potentials equivalent to that in a liquid sodium environment containing 2–5 molppm oxygen and 0.02–0.2 molppm carbon at temperatures 500–700 °C. Two He–CO–CO 2 environments (Pco/Pco 2 = 1320, Pco = 1980 molppm, and Pco/Pco 2 = 9, Pco = 13.5 molppm) were selected to test alloys NF616 and 316L at 700 and 850 °C. Upon exposure to He environments at 850 °C, 316L samples exhibited thick surface Cr 2 O 3 scales and substantial internal oxidation; however at 700 °C no significant internal oxidation was observed. NF616 samples exhibited relatively thinner surface Cr 2 O 3 scales compared to 316L samples at both temperatures. NF616 samples exposed to liquid sodium at 700 °C and He–Pco/Pco 2 = 9 at 850 °C showed decarburization. No surface oxide formation was observed on the sample exposed to the Na environment. Results obtained from He exposure experiments provide insight into what may occur during long exposure times in a sodium environment.

High-Cr (9 wt.%) ferritic–martensitic steels are important materials for use in nuclear reactors. This study shows a development activity for this category of steels via thermomechanical treatment (TMT) optimization and alloying element adjustment based on Grade 92 steels. Vickers microhardness and tensile tests were employed to assess the mechanical properties of the materials in the normalized-tempered (N&T) and optimized TMT conditions. The treatment of one of the modified heats produced ∼29% and ∼47% increases in hardness and yield strength, respectively, compared to the Grade 92 in the N&T condition. The TMT-treated alloys showed comparable or superior strength relative to the oxide-dispersion-strengthened steel PM2000. Microstructure analyses by optical and transmission electron microscopy together with thermodynamic calculations identified the strengthening mechanisms of the TMT and precipitates.

This work developed thermodynamic models for describing phase stability and thermodynamic property of the Fe–Cr–Zr system using the Calphad approach coupled with experimental study. Thermodynamic descriptions of the Fe–Cr and Cr–Zr systems were either directly adopted or slightly modified from literature. The Fe–Zr system has been remodeled to accommodate recent ab-initio calculation of formation enthalpies of various Fe–Zr compounds. Reliable ternary experimental data and thermodynamic models were mainly available in the Zr-rich region. Therefore, selected ternary alloys located in the vicinity of the eutectic valley of β(Fe,Cr,Zr) and (Fe,Cr)2Zr laves phase in the Fe-rich region have been experimentally investigated in this study. Microstructure has been examined by using scanning electron microscope, energy-dispersive X-ray spectroscopy and X-ray diffraction. These experimental results, along with the literature data were then used to develop thermodynamic models for phases in the Fe–Cr–Zr system. Calculated phase equilibria and thermodynamic properties of the ternary system yield satisfactory agreements with available experimental data.

Grain boundary engineering (GBE), primarily implemented by thermomechanical processing, is an effective and economical method of enhancing the properties of polycrystalline materials. Among the factors affecting grain boundary character distribution, literature data showed definitive effect of grain size and texture. GBE is more effective for austenitic stainless steels and Ni-base alloys compared to other structural materials of nuclear reactors, such as refractory metals, ferritic and ferritic–martensitic steels, and Zr alloys. GBE has shown beneficial effects on improving the strength, creep strength, and resistance to stress corrosion cracking and oxidation of austenitic stainless steels and Ni-base alloys.

An optimized thermomechanical treatment (TMT) applied to austenitic alloy 800H (Fe–21Cr–32Ni) had shown significant improvements in corrosion resistance and basic mechanical properties. This study examined its effect on radiation resistance by irradiating both the solution-annealed (SA) and TMT samples at 500 °C for 3 dpa. Microstructural characterization using transmission electron microscopy revealed that the radiation-induced Frank loops, voids, and γ′-Ni3(Ti,Al) precipitates had similar sizes between the SA and TMT samples. The amounts of radiation-induced defects and more significantly γ′ precipitates, however, were reduced in the TMT samples. These reductions would approximately reduce by 40.9% the radiation hardening compared to the SA samples. This study indicates that optimized-TMT is an economical approach for effective overall property improvements.

Ferritic/Martensitic (F/M) steels containing 9 wt.% Cr are candidates for structural and cladding components in the next generation of advanced nuclear fission and fusion reactors. Although it is known these alloys exhibit radiation-induced segregation (RIS) at grain boundaries (GBs) while in-service, little is known about the mechanism behind RIS in F/M steels. The classical understanding of RIS in F/M steels presents a mechanism where point defects migrate to GBs acting as perfect sinks. However, variation in grain boundary structure may influence the sink efficiency and these migration processes. A proton irradiated 9 wt.% Cr model alloy steel was investigated using STEM/EDS spectrum imaging and GB misorientation analysis to determine the role of GB structure on RIS at different GBs. An ab initio based rate theory model was developed and compared to the experimental findings. This investigation found Cr preferentially segregates to specific GB structures. The preferential segregation to specific GB structures suggests GB structure plays a key role in the mechanism behind radiation-induced segregation, showing that not all grain boundaries in F/M steels act as perfect sinks. The study also found how irradiation dose and temperature impact the radiation-induced segregation response in F/M steels.

Creep–fatigue testing of Ni-based superalloy Inconel 617 and Haynes 230 were conducted in the air at 850 °C. Tests were performed with fully reversed axial strain control at a total strain range of 0.5%, 1.0% or 1.5% and hold time at maximum tensile strain for 3, 10 or 30 min. In addition, two creep–fatigue life prediction methods, i.e. linear damage summation and frequency-modified tensile hysteresis energy modeling, were evaluated and compared with experimental results. Under all creep–fatigue tests, Haynes 230 performed better than Inconel 617. Compared to the low cycle fatigue life, the cycles to failure for both materials decreased under creep–fatigue test conditions. Longer hold time at maximum tensile strain would cause a further reduction in both material creep–fatigue life. The linear damage summation could predict the creep–fatigue life of Inconel 617 for limited test conditions, but considerably underestimated the creep–fatigue life of Haynes 230. In contrast, frequency-modified tensile hysteresis energy modeling showed promising creep–fatigue life prediction results for both materials.

This paper summarizes an investigation of the deformation hardening behavior of neutron-irradiated stainless steels and copper in terms of true stress(σ)–true strain(e) curves. It is commonly accepted that the σ–e curves are more informative for describing plastic flow, but there are few papers devoted to using the true curves for describing constitutive behavior of irradiated materials. This study uses previously published true and engineering curves for stainless steel and copper irradiated to different damage level. The most appropriate constitutive equation has been identified, and it is shown that for the strain range 0–0.6 the true curves can be well described by the Swift equation: σ = k(e − e0)0.5. The influence of irradiation on the parameters of the Swift equation is investigated in detail. It is found that in most cases the k-parameter of this equation is not changed significantly by irradiation. Since large data scatter was observed for the e0-parameter, a modified Swift equation σ = k(e − σ02/k2)0.5 was proposed and evaluated. This equation is based on the concept of an initial stress σ0, which is, in general, close to the yield stress. The relationships among k, e0, and damage dose, influence of test temperature and grain size are discussed.

Ferritic–martensitic (F–M) steels with 9 wt.%Cr are important structural materials for use in advanced nuclear reactors. Alloying composition adjustment, guided by computational thermodynamics, and thermomechanical treatment (TMT) were employed to develop high strength 9Cr F–M steels. Samples of four heats with controlled compositions were subjected to normalization and tempering (N&T) and TMT, respectively. Their mechanical properties were assessed by Vickers hardness and tensile testing. Ta-alloying showed significant strengthening effect. The TMT samples showed strength superior to the N&T samples with similar ductility. All the samples showed greater strength than NF616, which was either comparable to or greater than the literature data of the PM2000 oxide-dispersion-strengthened (ODS) steel at temperatures up to 650 °C without noticeable reduction in ductility. A variety of microstructural analyses together with computational thermodynamics provided rational interpretations on the strength enhancement. Creep tests are being initiated because the increased yield strength of the TMT samples is not able to deduce their long-term creep behavior.

Radiation-induced segregation in ferritic–martensitic alloy T 91 was studied to understand the behavior of solutes as a function of dose and temperature. Irradiations were conducted using 2 MeV protons to doses of 1, 3, 7 and 10 dpa at 400 °C. Radiation-induced segregation at prior austenite grain boundaries was measured, and various features of the irradiated microstructure were characterized, including grain boundary carbide coverage, the dislocation microstructure, radiation-induced precipitation and irradiation hardening. Results showed that Cr, Ni and Si segregate to prior austenite grain boundaries at low dose, but segregation ceases and redistribution occurs above 3 dpa. Grain boundary carbide coverage mirrors radiation-induced segregation. Irradiation induces formation of Ni–Si–Mn and Cu-rich precipitates that account for the majority of irradiation hardening. Radiation-induced segregation behavior is likely linked to the evolution of the precipitate and dislocation microstructures.

Casting of austenitic stainless steels offers the possibility of directly producing large and/or relatively complex structures, such as the first wall shield modules or the divertor cassette for the ITER, which may lead to simpler component fabrication and major cost savings. Past efforts to use cast steel for these large components were unsuccessful due to lower than acceptable strength in the test components. To improve and validate cast stainless steel as a substitute for wrought stainless steel for shield module applications, a series of test cast steels based on the commercially available CF3M specification have been designed and fabricated. These modifications utilize combinations of Mn and N, which result in significant increases in strength, fracture toughness, and impact properties. These mechanical performance improvements have been achieved without any loss of irradiation performance, corrosion performance, or weldability.

Ferritic–martensitic (F/M) alloys are expected to play an important role as cladding or structural components in Generation IV and other advanced nuclear systems operating in the temperature range 350–700 °C and to doses up to 200 displacements per atom (dpa). Oxide dispersion strengthened (ODS) F/M steels have been developed to operate at higher temperatures than traditional F/M steels. These steels contain nanometer-sized Y–Ti–O nanoclusters for additional strengthening. A proton irradiation to 1 dpa at 525 °C has been performed on a 9Cr ODS steel to determine the nanocluster stability at low dose. The evolution of the nanocluster population and the composition at the nanocluster–matrix interface were studied using electron microscopy and atom probe tomography. The data from this study are contrasted to those from a previous study on heavy-ion irradiated 9Cr ODS steel.

Thermophysical and mechanical properties of high purity chemically vapor-deposited (CVD) SiC and chemically vapor-infiltrated SiC matrix, pyrocarbon/SiC multilayered interphase composites with Hi-Nicalon™ Type-S and Tyranno™-SA3 SiC fibers were evaluated following neutron irradiation. Specimens including statistically significant population of tensile bars were irradiated up to 5.3 displacement-per-atom at ∼220 to ∼1080 °C in the Advanced Test Reactor at Idaho National Laboratory and High Flux Isotope Reactor at Oak Ridge National Laboratory. Thermal diffusivity/conductivity of all materials decreased during irradiation. The reciprocal thermal diffusivity linearly increased with temperature from ambient to the irradiation temperature. The magnitude of defect thermal resistance was distinctively different among materials and its ranking was Hi-Nicalon™ Type-S > Tyranno™-SA3 > CVD SiC regardless of irradiation condition. Dynamic Young’s modulus decrease for the irradiated CVD SiC exhibited explicit correlation with swelling. No significant effects of neutron irradiation on tensile properties of the composites were revealed, except for an anomaly case for the Hi-Nicalon™ Type-S composite irradiated in a specific condition. According to the single filament tensile evaluation, fibers of both types retained the original strength during irradiation at intermediate temperatures but significantly deteriorated during bare fiber irradiation at ∼910 °C. However, fiber strength deterioration was not observed when irradiated in composite form. Irradiation effects on the fiber–matrix interface properties were discussed based on results from the composite and single filament tensile tests, the hysteresis analysis, and the fracture surface examination.

A key obstacle to the commercial deployment of advanced fast reactors is the capital cost. There is a perception of higher capital cost for fast reactor systems than advanced light water reactors. However, cost estimates come with a large uncertainty since far fewer fast reactors have been built than light water reactor facilities. Furthermore, the large variability of industrial cost estimates complicates accurate comparisons. Reductions in capital cost can result from design simplifications, new technologies that allow reduced capital costs, and simulation techniques that help optimize system design. It is plausible that improved materials will provide opportunities for both simplified design and reduced capital cost. Advanced materials may also allow improved safety and longer component lifetimes. This work examines the potential impact of advanced materials on the capital investment cost of fast nuclear reactors.

Zirconium or hafnium additions to austenitic stainless steels caused a reduction in grain boundary Cr depletion after proton irradiations for up to 3 dpa at 400 °C and 1 dpa at 500 °C. The predictions of a radiation-induced segregation (RIS) model were also consistent with experiments in showing greater effectiveness of Zr relative to Hf due to a larger binding energy. However, the experiments showed that the effectiveness of the solute additions disappeared above 3 dpa at 400 °C and above 1 dpa at 500 °C. The loss of solute effectiveness with increasing dose is attributed to a reduction in the amount of oversized solute from the matrix due to growth of carbide precipitates. Atom probe tomography measurements indicated a reduction in amount of oversized solute in solution as a function of irradiation dose. The observations were supported by diffusion analysis suggesting that significant solute diffusion by the vacancy flux to precipitate surfaces occurs on the time scales of proton irradiations. With a decrease in available solute in solution, improved agreement between the predictions of the RIS model and measurements were consistent with the solute-vacancy trapping process, as the mechanism for enhanced recombination and suppression of RIS.

Ferritic–martensitic (FM) alloys are expected to play an important role as cladding or structural components in Generation IV systems operating in the temperature range 350–700 °C and to doses up to 200 dpa. Oxide dispersion strengthened (ODS) ferritic–martensitic steels have been developed to operate at higher temperatures than traditional FM steels. These steels contain nanometer-sized Y–Ti–O nanoclusters as a strengthening mechanism. Heavy ion irradiation has been used to determine the nanocluster stability over a temperature range of 500–700 °C to doses of 150 dpa. At all temperatures, the average nanocluster size decreases but the nanocluster density increases. The increased density of smaller nanoclusters under radiation should lead to strengthening of the matrix. While a reduction in size under irradiation has been reported in some other studies, many report oxide stability. The data from this study are contrasted to the available literature to highlight the differences in the reported radiation response.

It is the general conclusion of all national programs that the development of high-performance reduced-activation structural materials is essential for the successful development of fusion power. In this paper, the experience gleaned from previous programs to develop materials for high temperature structural applications is used to identify and discuss some of the most critical issues that must be addressed in the development of candidate materials for fusion structural applications. Critical issues discussed include radiation-induced solute segregation and implications on phase stability in the development of high-performance alloys/ceramics; the effects of very large amounts of helium on mechanical properties and the implications for alloy design/development; development of high temperature design methodology and incorporation of radiation effects into this methodology; the effects of radiation damage on flow localization, and the implications and approach to control the phenomena; and considerations of mass transfer and corrosion in complex fusion systems.

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.

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.

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.

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.

The deformation microstructure of proton-irradiated stainless steels may play a key role in explaining their irradiation-assisted stress corrosion cracking (IASCC) susceptibility. In the present study, three model alloys (UHP-304, 304 + Si, 304 + Cr + Ni) with different stacking fault energies (SFEs) were irradiated with 3.2 MeV protons at 360 °C to 1.0 and 5.5 dpa and then strained in 288 °C Ar atmosphere. The deformation microstructure of the strained samples was investigated using scanning electron microscopy and transmission electron microscopy. The results showed that the slip lines interacted with grain boundaries by grain-to-grain transmission, grain boundary sliding or deformation ledge formation at grain boundaries. Expanded channels, which were formed at locations where dislocation channels intersected the grain boundaries or other channels, were found predominately in the low SFE alloys UHP-304 and 304 + Si. The steps and shear strain at grain boundaries caused by channel expansion may increase the IASCC susceptibility in low SFE stainless steels by producing strain concentrations and inducing cracks in the oxide film.