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Neutron irradiation in mixed spectrum reactors is an effective approach to investigate the synergistic effects of solid transmutations and atomic displacement damage in tungsten. In this article the findings and interpretations from a large-scale tungsten irradiation campaign conducted using the High Flux Isotope Reactor of Oak Ridge National Laboratory are summarized. The response of originally-unalloyed tungsten to mixed spectrum neutron irradiation is characterized by enormous hardening, degradation of modulus of toughness eventually leading to embrittlement, and decreased thermal conductivity. These property changes may be attributed primarily to production, segregation, and precipitation of rhenium and osmium and assisted by displacement damage. The stages describing the evolution of the microstructural development and property changes with the neutron dose and the accompanying accumulation of transmutation products are proposed. This manuscript has been co-authored by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan ( http://energy.gov/downloads/doe-public-access-plan ).

Thirteen austenitic stainless steels, nickel-base alloys, and ferritic alloys were irradiated using 2 MeV protons at 360 °C to a damage level of 2.5 displacements per atom (dpa). Comprehensive microstructural characterization was performed for irradiation-induced features, including dislocation loops, voids, precipitates, and radiation induced segregation (RIS). Dislocation loops formed in all alloys except 14YWT, while voids were observed in alloys 316 L, 310, C22, and 14YWT. Irradiation-induced formation of γ′ precipitates was observed in alloys 316 L, 310, 800, and 690; the irradiation-enhanced, long-range ordered Ni2Cr phase (Pt2Mo-type) was observed in alloys 690, C22, 625, 625Plus, 625DA, and 725; and G-phase was observed in alloy T92. No irradiation-induced precipitates were observed in alloys X750, 718 or 14YWT. Precipitation of the γ′ phase can be understood through segregation and clustering of Si, Al, and Ti. Overall, austenitic stainless steels are generally susceptible to irradiation damage in the form of loops, voids, precipitates, and RIS. Ni-base alloys have this same type of dislocation loops and RIS behaviors but are more resistant to void swelling. Ferritic alloys showed better resistance to loop formation, void swelling and irradiation-induced precipitation. From the degree of irradiation-induced microstructural change, alloy T92 was identified as the most radiation resistant among these alloys.

Tungsten is the primary candidate for plasma-facing materials (PFMs) in a magnetic fusion energy (MFE) devices such as ITER due to its high melting point, excellent erosion resistance, and low sputtering yield. However, tungsten will suffer from heat, neutron irradiation, large flux (1022-1024 He/m2s)- low energy (tens of eV to hundreds of eV) helium and hydrogen ion exposure, as well as microstructure evolution (such as recrystallization). We have exposed hot-rolled and recrystallized tungsten to 65 or 80 eV helium ions with a flux of 0.5 × 1022 or 5 × 1022 He/m2s to total fluence from 0.6 × 1024 to 4 × 1024 He/m2 at a temperature of ≈1100 K. The results show that recrystallized tungsten samples exhibit a roughened surface morphology in certain grains with orientations close to , while hot-rolled tungsten samples still maintain the smooth surface under all helium-ion exposure conditions. All the samples exhibit relatively shallow helium bubble penetration (

Modern electron microscopy techniques were used to conduct a thorough study of an irradiated urania fuel pellet microstructure to attempt at an understanding of high burnup structure formation in this material. The fuel was irradiated at low power to high burnups in a light water reactor, proving ideal for this purpose. Examination of grain size and orientation with strict spatial selectivity across the fuel pellet radius allowed for capturing the progression of the restructuring process, from its onset to full completion. Based on this information, the polygonization mechanism was shown to be responsible for restructuring, involving formation of low-angle grain boundaries with their initiation occurring at the original high-angle grain boundaries of the as-fabricated pellet and at the gas bubble-matrix interfaces. The low-angle character of boundaries between the subdivided grains disappeared in the fully developed high burnup structure, likely due to creep deformation in the pellet.

Tungsten is the primary candidate for plasma-facing materials in the critical regions of a magnetic fusion energy (MFE) device, such as a tokamak. However, tungsten (and many other metals) grow copious nanotendril "fuzz" when exposed to fusion-relevant He-bearing plasmas, and the impact of this fuzz on reactor operation is still unknown. Further, the mechanism of fuzz growth is also poorly understood at present. In order to experimentally probe the growth mechanisms, we developed a new sample preparation technique to examine the interface of individual nanotendril roots with the substrate (substrate/tendril interface) using electron microscopy and measured the grain boundary character distributions (GBCDs) of substrate/tendril interfaces, the GBs within individual tendrils (tendril/tendril interfaces), and GBs within the substrate (substrate/substrate interfaces). We find the substrate/tendril and tendril/tendril GBCDs are essentially indistinguishable, but very different from the substrate/substrate GBCDs, which implies that tendril growth periodically forms new GBs at the substrate/tendril interface, and these GBs are pushed upward to become tendril/tendril interfaces within the fuzz mat. These results will help guide future computational modelling studies to elucidate the details of the growth mechanisms.

Three types of microstructures, i.e., tempered-martensite (TM), ferrite (F), and dual-phase (TM + F), were developed in a castable nanostructured alloy that favors a high density of nanoprecipitates compared with the precipitates in current reduced-activation ferritic-martensitic steels. The effect of the distinct microstructures on tensile properties, Charpy impact toughness, and thermal helium desorption behavior was investigated with the full TM structure as a reference. The results indicated that the F domain in the TM + F structure governed the strength and slightly impaired the impact toughness. The full F structure exhibited the highest strength without compromising ductility, but it noticeably diminished impact toughness. All microstructures had a dominant helium desorption peak at ∼1070 °C. The higher density of nanoprecipitates and complex boundaries and dislocations in the TM + F structure enhanced the secondary helium desorption peak and extended the shoulder peak, in contrast to the full TM structure with an enlarged desorption peak associated with the ferrite-to-austenite transformation at ∼810–850 °C and the full F structure with a dominant desorption peak related to bubble migration at ∼1070 °C. These results suggest that components fabricated from functionally graded microstructures could be engineered to exploit the advantages of different microstructures for demanding application requirements.

A reduced activation ferritic/martensitic steel, Eurofer97, was neutron irradiated in the vicinity of 300 °C in the High Flux Isotope Reactor (HFIR) up to 72 dpa. Advanced analytical scanning transmission electron microscopy and conventional transmission electron microscopy were applied to investigate the radiation-induced segregation and phase instability behavior after neutron irradiation. Amorphization was observed in M23C6 carbides. Cr-rich clusters were seen within the matrix, near the lath boundaries and close to the M23C6 carbides. Cr enrichment and Fe depletion were detected at both prior austenite grain boundaries and lath boundaries, despite different segregation magnitude. In addition, the enrichment of Ni, the depletion of V, and tiny cavities (presumably helium bubbles) are also found at lath boundaries. This work interrogates the evolution of microstructures after neutron irradiation, which provides detailed understanding on the microstructural aspects controlling the mechanical integrity of Eurofer97 under high-dose neutron damage.

Graphite is the moderator material of several Generation IV nuclear reactor concepts, as well as the British Advanced Gas-cooled Reactors (AGR). Porosity can heavily influence the material properties, mechanical irradiation response, and neutron induced shrinkage or swelling of nuclear-grade graphite. Due to the sub-micron size of several types of pores found in graphite, only a high-resolution imaging technique such as electron tomography are capable of visualizing these features in three dimensions. In this research, we used electron tomography to characterize as-received and neutron irradiated samples of IG-110 nuclear-grade graphite to show for the first time the 3D structure of both native and irradiation-induced nano-cracks. This technique also reveals unique characteristics of graphite such as the structure that surrounds pores and could be used to inform molecular dynamic simulations of irradiated graphite and experimental techniques such as gas-absorption. This research also shows the utility of this technique for the study of other nuclear porous carbon-based materials.

Understanding the effects of helium on microstructures and mechanical properties of reduced-activation ferritic-martensitic steels is important to use of these steels in fusion reactor structures. The 9Cr-2WVTa steels were doped with 58Ni and 60Ni isotopes at 2 weight percent to control the rate of transmutation helium generation. The samples were irradiated in the High Flux Isotope Reactor to ~24 displacements per atom at nominal temperatures of 300, 400, and 500°C, producing 228 and 7 atomic parts-per-million helium in the 58Ni- and 60Ni-doped samples, respectively. Transmission electron microscopy revealed a variety of precipitates and the radiation-induced dislocation loops and cavities (voids or helium bubbles). Tensile tests of the irradiated samples at the irradiation temperatures showed radiation-induced hardening at 300°C and radiation-induced softening at 400°C. Analysis indicates that the hardening primarily originated from the loops and cavities. The 58Ni-doped samples had greater strengthening contributions from loops and cavities, leading to higher hardening with lower ductility than the 60Ni-doped samples. The greater helium production of 58Ni did not show pronounced reductions in ductility of the samples.

Nanostructured ferritic alloys (NFAs) are new generation materials for use in high temperature energy systems, such as nuclear fission or fusion reactors. However, joining these materials is a concern, as their unique microstructure is destroyed by traditional liquid-state welding methods. The microstructural evolution of a friction stir welded 14YWT NFA was investigated by atom probe tomography, before and after a post-weld heat treatment (PWHT) at 1123K. The particle size, number density, elemental composition, and morphology of the titanium-yttrium-oxygen-enriched nanoclusters (NCs) in the stir and thermally-affected zones were studied and compared with the base metal. No statistical difference in the size of the NCs was observed in any of these conditions. After the PWHT, increases in the number density and the oxygen enrichment in the NCs were observed. Therefore, these new results provide additional supporting evidence that friction stir welding appears to be a viable joining technique for NFAs, as the microstructural parameters of the NCs are not strongly affected, in contrast to traditional welding techniques.

Characterization methods capable of providing critical information across multiple structural length scales are essential in materials exposed to the extreme environments such as anticipated fusion power systems. Complementary techniques capable of uncovering the complicated microstructural irradiation-induced evolution are also important to verify and validate advanced computational models. To date, the primary microstructural tools informing such lower-length scale models have included analytical electron microscopy, positron annihilation spectroscopy, atom probe tomography, and small-angle neutron scattering. In this paper, we discuss the application of state-of-the-art synchrotron-based x-ray characterization methods in fusion material research. Specifically highlighted are opportunities in leveraging synchrotron-based techniques to address fundamental and applied materials science challenges at various length scales and in support of modeling efforts. Examples presented in this article include: a combined small angle x-ray scattering and x-ray diffraction study of transmutation-induced precipitation in neutron irradiated tungsten, and the identification of size and structure of nm-scale transmutation precipitates and voids; quantitative characterization of thermodynamically predicted minor precipitate populations in advanced reduced activation ferritic-martensitic steels through high energy x-ray diffraction; and a review of recent synchrotron-based studies dedicated to quantifying the radiation response of fusion relevant materials. The latter includes a pair distribution function analysis investigation of neutron irradiated SiC with insights into the different radiation response of the silicon and carbon sublattices, and a dose dependent decrease in the size of defect free material.

Nanostructured ferritic alloys (NFAs) have outstanding high temperature creep properties and extreme tolerance to radiation damage. To achieve these properties, NFAs are fabricated by mechanical alloying of metallic and yttria powders. Atom probe tomography has demonstrated that milling times of at least 40 h are required to produce a uniform distribution of solutes in the flakes. After milling and hot extrusion, the microstructure consists of -Fe, high number densities of Ti-Y-O-vacancy-enriched nanoclusters, and coarse Y2Ti2O7 and Ti(O,C,N) precipitates on the grain boundaries. In contrast, the as-cast condition consists of -Fe with 50-100 m irregularly-shaped Y2Ti2O7 pyrochlore precipitates with smaller embedded precipitates with the Al5Y3O12 (yttrium-aluminum garnet) crystal structure indicating that this traditional processing route is not a viable approach to achieve the desired microstructure. The nano-hardnesses were also substantially different, i.e., 4 and 8 GPa for the as-cast and as-extruded conditions, respectively. These differences can be explained by the differences in the microstructure and the effects of the high vacancy content introduced by mechanical alloying, and the strong binding energy of vacancies with O, Ti, and Y atoms retarding diffusion.

Diffusion plays a major role in the stability of microstructures to extreme conditions of high temperature and high doses of irradiation. In nanostructured ferritic alloys, first principle calculations indicate that the binding energy of vacancies is reduced by the presence of oxygen, titanium and yttrium atoms. Therefore, the number of free vacancies available for diffusion can be greatly reduced. The mechanical properties of these alloys, compared to traditional wrought alloys of similar composition and grain structure, is distinctly different, and the ultrafine grained alloy is distinguished by a high number density of Ti–Y–O-enriched nanoclusters and solute clusters, which drives the mechanical response. When a displacement cascade interacts with a nanocluster, the solute atoms are locally dispersed into the matrix by ballistic collisions, but immediately a new nanocluster reforms due to the local supersaturation of solutes and vacancies until the excess vacancies are consumed. The result of these processes is a structural material for advanced energy systems with a microstructure that is self-healing and tolerant to high doses of radiation and high temperatures.

Nanostructured ferritic alloys (NFAs) exhibit complex microstructures consisting of 100–500 nm ferrite grains, grain boundary solute enrichment, and multiple populations of precipitates and nanoclusters (NCs). Understanding these materials’ excellent creep and radiation-tolerance properties requires a combination of multiple atomic-scale experimental techniques. Recent advances in scanning transmission electron microscopy (STEM) hardware and data analysis methods have the potential to revolutionize nanometer-to micrometer-scale materials analysis. Modern high-brightness, high-X-ray collection STEM instruments are capable of enabling advanced experiments, such as simultaneous energy dispersive X-ray spectroscopy and electron energy loss spectroscopy spectrum imaging at nm to sub-nm resolution, that are now well-established for the study of nuclear materials. In this paper, we review past results and present new results illustrating the effectiveness of latest-generation STEM instrumentation and data analysis.

A nondestructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless-steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Our study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount (≈20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor can be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for determining absolute hydrogen concentrations.

An effort is underway at the Spallation Neutron Source (SNS) to redesign the 1st Generation high-nickel alloy proton beam window (PBW) to a 2nd Generation design that utilizes aluminum for the window material. One of the key challenges to implementation of an aluminum PBW at the SNS was selection of an appropriate joining method to bond an aluminum window to the stainless steel shielding structure of the PBW assembly. An explosively formed bond was selected as the most promising joining method for the aluminum PBW design and a testing campaign was conducted to evaluate the strength and efficacy of explosively formed bonds that were produced using two different interlayer materials: niobium and titanium. The characterization methods reported here include tensile testing, thermal-shock leak testing, optical microscopy, and advanced scanning electron microscopy. All tensile specimens examined failed in an aluminum interlayer and measured tensile strengths were all slightly greater than the native properties of the aluminum interlayer, while elongation values were all slightly lower. A leak developed in the test vessel with an niobium interlayer after repeated thermal-shock cycles, which was attributed to an extensive crack network that formed in an interfacial layer of a niobium-rich constituent phase located on the bond interfaces of the niobium interlayer; the test vessel with a titanium interlayer did not develop a leak under the conditions tested. Due to the experience gained from these characterizations, an explosively formed bond with a titanium interlayer was selected for the aluminum PBW design at the SNS.

The response of a nanostructured ferritic alloy to He implantation and post-irradiation annealing (PIA) at 750 °C was characterized by atom probe tomography and transmission electron microscopy. The supersaturated He concentration in the ferrite at a dose of ∼2.1 displacements per atom was similar for the as-implanted, 75 ± 7 appm, and a 10 h PIA treatment, 71 ± 7 appm, but decreased to 38 ± 2 appm after a 100 h PIA treatment. Approximately 91–97% of the He bubbles were present as isolated bubbles in the ferrite and ∼1–5% on the surface of the nanoclusters in the ferrite. The remainder were associated with the grain boundaries with a small fraction on the surface of Ti(N,C,O) precipitates. Their average size and number density were similar for the as-implanted and 10 h PIA treatment with a small increase in the size and a significant increase in the number density after the 100 h PIA treatment. Swelling in the high dose region increased from ∼1% in the as-implanted and 10 h PIA conditions to ∼5% after the 100 h PIA treatment but the estimated number of He atoms per unit volume in the He bubbles decreased by an order of magnitude. Number densities increased from ∼8 × 10 23 m −3 in the as-implanted to ∼15 × 10 23 m −3 in the 10 h PIA condition, with little change (to ∼12 × 10 23 m −3 ) in the 100 h PIA condition. This trend may indicate nucleation of new bubbles up to 10 h, with growth and possible consumption of the smaller bubbles between 10 and 100 h.

A latest-generation aberration-corrected scanning/transmission electron microscope (STEM) is used to study heavy-ion-irradiated nanostructured ferritic alloys (NFAs). Results are presented for STEM X-ray mapping of NFA 14YWT irradiated with 10 MeV Pt to 16 or 160 dpa at −100 °C and 750 °C, as well as pre-irradiation reference material. Irradiation at −100 °C results in ballistic destruction of the beneficial microstructural features present in the pre-irradiated reference material, such as Ti–Y–O nanoclusters (NCs) and grain boundary (GB) segregation. Irradiation at 750 °C retains these beneficial features, but indicates some coarsening of the NCs, diffusion of Al to the NCs, and a reduction of the Cr–W GB segregation (or solute excess) content. Ion irradiation combined with the latest-generation STEM hardware allows for rapid screening of fusion candidate materials and improved understanding of irradiation-induced microstructural changes in NFAs.

A 14YWT nanostructured ferritic alloy has been irradiated with 335 keV He+ to a total fluence of 6.75 × 1020 He m−2 at a temperature of 400 °C and subsequently thermally treated at 750 °C for up to 100 h. Transmission electron microscopy has been used to characterize the size and distribution of the resultant helium bubbles. The results indicate that the bubbles generally increase in size and the distribution becomes more inhomogeneous during the thermal treatment. The results are discussed in terms of the helium supply and vacancy supersaturation, Brownian motion and coalescence, and Ostwald ripening mechanisms.

Ferritic–martensitic (FM) steels have been widely considered as structural materials for future fusion and fission systems due to their excellent swelling resistance and other properties under irradiation. NF616 is a newer generation FM steel that provides higher allowable operating temperatures, but there is limited information on the development of irradiated microstructures or associated mechanical properties in NF616. Studying the irradiation-induced defects as a function of dose could help to better understand the mechanisms of damage evolution. Proton irradiation at 773 K, up to 1, 2 and 3 dpa, was performed on NF616 to investigate damage evolution at different doses. Irradiation-induced defects, such as clusters, loops and dislocations, were investigated.

The viability of advanced oxidation-resistant Fe-base alloys to protect zirconium from rapid oxidation in high-temperature steam environments has been examined. Specimens were produced such that outer layers of FeCrAl ferritic alloy and Type 310 austenitic stainless steel were incorporated on the surface of zirconium metal slugs. The specimens were exposed to high-temperature 0.34 MPa steam at 1200 and 1300 °C. The primary degradation mechanism for the protective layer was interdiffusion with the zirconium, as opposed to high-temperature oxidation in steam. The FeCrAl layer experienced less degradation and protected the zirconium at 1300 °C for 8 h. Constituents of the Fe-base alloys rapidly diffused into the zirconium and resulted in the formation of various intermetallic layers at the interface and precipitates inside the bulk zirconium. The nature of this interaction for FeCrAl and 310SS has been characterized by use of microscopic techniques as well as computational thermodynamics. Finally, a reactor physics discussion on the applicability of these protective layers in light-water-reactor nuclear fuel structures is offered.

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.

Stress relaxation of elastically strained silicon carbide samples during high flux neutron irradiation to ∼2 displacements per atom at intermediate (390–540 °C) to high (790–1180 °C) temperatures is presented. The magnitude of stress relaxation normalized to the initial stress magnitude is independent of the initial stress magnitude, indicating a stress exponent of unity for irradiation creep in SiC. The creep strain increases with the increasing fluence while the strain rate significantly decreases. A linear relationship was found between the creep strain and the transient swelling that occurs due to irradiation defect accumulation. The apparent irradiation creep compliances for silicon carbide are substantially smaller than those associated with pure metals and alloys. Microstructural examination suggests that incoherent grain boundaries likely play a major role in determining the primary transient irradiation creep of these materials at high temperatures with a potential additional contribution from basal slip at very high temperatures.

A 14YWT nanostructured ferritic alloy (NFA) was implanted with He + ions to fluences of 6.75 × 1021 He m-2 at 400 °C in order to simulate the effects of high He concentrations produced in advanced ...

The current generation of fully ceramic microencapsulated fuels, consisting of Tristructural Isotropic fuel particles embedded in a silicon carbide matrix, is fabricated by hot pressing. Matrix powder feedstock is comprised of alumina–yttria additives thoroughly mixed with silicon carbide nanopowder using polyethyleneimine as a dispersing agent. Fuel compacts are fabricated by hot pressing the powder–fuel particle mixture at a temperature of 1800–1900 °C using compaction pressures of 10–20 MPa. Detailed microstructural characterization of the final fuel compacts shows that oxide additives are limited in extent and are distributed uniformly at silicon carbide grain boundaries, at triple joints between silicon carbide grains, and at the fuel particle–matrix interface.

Erbium metal thin-films have been deposited on molybdenum-on-silicon substrates and then converted to erbium dideuteride (ErD{sub 2}). Here, we study the effects of deposition temperature ({approx}300 or 723 K) and deposition rate (1 or 20 nm/s) upon the initial Er metal microstructure and subsequent ErD{sub 2} microstructure. We find that low deposition temperature and low deposition rate lead to small Er metal grain sizes, and high deposition temperature and deposition rate led to larger Er metal grain sizes, consistent with published models of metal thin-film growth. ErD{sub 2} grain sizes are strongly influenced by the prior-metal grain size, with small metal grains leading to large ErD{sub 2} grains. A novel sample preparation technique for electron backscatter diffraction of air-sensitive ErD{sub 2} was developed, and allowed the quantitative measurement of ErD{sub 2} grain size and crystallographic texture. Finer-grained ErD{sub 2} showed a strong (1 1 1) fiber texture, whereas larger grained ErD{sub 2} had only weak texture. We hypothesize that this inverse correlation may arise from improved hydrogen diffusion kinetics in the more defective fine-grained metal structure or due to improved nucleation in the textured large-grain Er.