The plasma-facing components (PFCs) of future fusion reactors will have intricate structures and require multiple materials because no one material can simultaneously satisfy all the requirements of the component. The dissimilar material joints in PFCs must withstand extreme thermal and stress gradients under neutron irradiation. The Fusion Research Oriented to Neutron Irradiation and Tritium Behavior at Material Interfaces (FRONTIER) U.S.-Japan collaboration seeks to explore and explain the behavior of internal solid interfaces in PFCs under neutron irradiation. The first step of the collaboration was to identify the leading PFCs that should be studied further and prepare them for the next step, which will include neutron irradiation. Different strategies for material development are being pursued worldwide to produce robust PFCs. Here, an overview is presented of some of the most promising materials in the areas of copper alloys, tungsten-copper composites, tungsten-steel composites, additively manufactured tungsten, particle-reinforced tungsten, and tungsten and SiC fiber composites. Each material's fabrication and benefits are described, and some discussion of remaining questions is given.

A major challenge for heat transfer in nuclear materials is to ensure thermal mobility after high amounts of neutron irradiation. Tungsten is widely selected as a heat transfer material in fusion reactors. In metals, thermal conductivity is dominated by electrons' ability to transfer energy. Neutron irradiation generates point defects, clusters, and solid transmutation (e.g.rhenium and osmium in tungsten), which inhibit electron motion. The purpose of this work is to quantify the irradiation-induced change in electron mobility and deconvolute transmutation and microstructural effects on observed changes to electron mobility. Single and polycrystalline tungsten were fast neutron irradiated in the High Flux Isotope Reactor at Oak Ridge National Laboratory to doses between 0.2 and 0.7 displacements per atom (dpa) and temperatures from 500 degrees C to 1000 degrees C. Grain growth was observed in all samples. Microstructure and transmutation were quantified. The geometric orientation of samples with elongated grains has been shown to affect electrical resistivity. A mathematical model was developed and used to deconvolute solid-solution transmutation, grain, and temperature-dependent lattice effects on resistivity. At similar to 0.4 dpa at similar to 590 degrees C, the combined resistivity degradation due to voids, vacancies, interstitials, and dislocations is estimated to be greater than the contribution from solid solution Re transmutation, which is greater than the contribution from grain boundaries. At doses of similar to 0.7 dpa at similar to 750 degrees C, solid solution Re contributions are greater than all other effects combined. This work establishes a basis to predict the effects of irradiation temperature and transmutation on thermal properties of tungsten and highlights the importance of irradiation temperature.

original material under consideration for the TSV was Zircaloy-4 (Zry-4), but it is known to form a hydride which can degrade its mechanical properties. Thus, some investigation here is focused on the effects of hydrogen uptake in the Zry-4. Additionally, the alternative material being considered is AISI 347, a stainless steel. Both materials have little existing data for their neutron irradiation behavior below 100°C, and both have open questions on the weld behavior under neutron irradiation. Testing has focused on characterizing their weld properties with tensile tests and performing neutron irradiation of samples in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL).

Tungsten is a candidate divertor material for many tokamak reactors and has potential to be durable to high heat flux conditions. However, plasma-materials interaction and edge localized modes (1-10 GW/m(2)) can cause significant hardening and melting damage. Tungsten in a high heat flux (46.3 GW/m(2)) and helium plasma environment was investigated. We studied possible residual helium and microstructure de-formation in resolidified tungsten. Following this, a 1-30 keV Ga focused ion beam was used for TEM sample milling. In as-received tungsten, dark spots of possible lattice strain and defects were in the grains. After high heat exposure under plasma pressure, intergranular features, dark spots in TEM, diffraction patterns, and elongated FIB induced pore-artifact structures (second phase) formed. Helium appears to be absent from the resolidified tungsten matrix due to erosive melting damage and high temperature conditions. A simple four-step microstructure deformation process from the elastic bulk side to the near-pore environment is proposed. The presence of defects in TEM images, grain size reduction, and microstructure deformation affected hardening. The near-pore environment likely experienced grain refinement, and further disordered nanophases are possible under severe deformation conditions. (C) 2022 Elsevier B.V. All rights reserved.

Digital holography has been proposed to fulfill a need for an imaging diagnostic capable of in situ monitoring of surface erosion caused by plasma-material interaction in nuclear fusion devices. A digital holography diagnostic for 3D surface erosion measurement has been developed at Oak Ridge National Laboratory with the goal of deployment on a plasma device. A proof-of-concept in situ demonstration is planned which would involve measurement of plasma erosion on targets exposed to an electrothermal arc source. This work presents the results of an ex situ characterization of the capability and limitations of holographic imaging of targets exposed to the arc source. Targets were designed to provide a fiducial for comparison of deformed and unaffected areas. The results indicated that the average net erosion was ∼150 nm/plasma exposure, which is expected to be within the diagnostic's measurement capacity. Surface roughness averages determined by holographic image analysis showed good agreement with measurements taken with a profilometer. The limit of the holography diagnostic's x-y spatial resolution was characterized by comparison with scanning electron microscope imaging.

An electrothermal (ET) capillary arc discharge is an effective source of high-density plasmas which can be used to evaluate high heat flux plasma impact on materials 1 . In this work, W and Mo samples were exposed to plasmas with densities between 10 19 and 10 23 m -3 , temperatures between 0.1 and 1 eV, and a mixed material composition from W/polycarbonate and Mo/polycarbonate liners. These plasma-impacted materials have been characterized using EDS, XPS, FIB, nanoindentation and refractometry in order to determine microstructure of the surface and bulk material, surface composition and sample wear behavior (strain behavior, surface roughness). The surface was confirmed to be primarily composed of metal-carbides on self-similar substrates, with trace amounts of other metals including copper from the cathode. Deposited materials showed good adhesion, with textured amorphous surfaces and grain sizes of about 50 nm. The penetration depth of ablated material and mechanical properties of the samples will be reported in detail, with early results suggesting nucleation is the primary mechanism behind surface structural organization during plasma deposition.

In support of the reactor-vessel development effort of SHINE medical technologies, irradiation of its structural materials and post-irradiation testing have been performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. This irradiation was performed on welded and unwelded AISI 347 steel and Zircaloy-4 samples at low temperature (~60°C) in contact with the reactor coolant. Hardness, Charpy, and tensile data have been taken from all cases of unirradiated specimens, while post-irradiation tensile and hardness data have been taken from AISI 347 base metal and Zircaloy-4 base metal and weld specimens so far. Additionally, 3-point bend data have been taken from explosion-welded specimens. The expected changes to hardness, yield strength, ultimate tensile strength, uniform elongation, and total elongation as a result of the low temperature and low dose neutron irradiation are reported here.

Opportunities exist to expand the application of Zircaloy-4 (Zry-4) into lower-temperature (<100°C) applications. Welded areas within these structures are potential failure points and post-weld heat treatments (PWHTs) are not yet standardized for application in this temperature range. Zry-4 tungsten inert gas (TIG) welds were given PWHTs from 450-900°C with hold times from 0.5-48 hours to investigate changes to tensile properties and microstructure. Treatment of 800°C for 1 hour was found to produce the most ductile material, measured by total elongation of the tensile bar (16.5%). Both PWHT time and temperature were observed to increase total elongation, except in extreme (>800°C or >18hours) cases, but the effect of temperature was stronger. The second order cumulative annealing parameter shows a trend with tensile properties. Average grain sizes in the fusion zone and heat affected zone did not change under any of the PWHTs until the development of an undesirable blocky-alpha phase, which was only observed to form in the heat affected zones for hold times greater than 18hr at 800°C.

The tensile properties of powder-metallurgical-processed Pure W, K-doped W, W-3%Re, and K-doped W-3%Re were examined after neutron irradiation up to 0.7 dpa at 910-1020 degrees C with a thermal neutron shield in the High Flux Isotope Reactor (HFIR). After irradiation, recrystallized Pure W (R) exhibited a brittle fracture mode, while recrystallized K-doped W-3%Re (R) exhibited a ductile fracture mode at 500 degrees C. K-doped W-3%Re (R) has fine grains, and hence, contains a considerable number of grain boundaries that act as sinks for irradiation defects. Solid solute Re in the W matrix could improve not only the mechanical properties of W, but also its resistance to neutron irradiation. At 500 degrees C, the ductility of K-doped W-3%Re after irradiation was significantly higher than that of Pure W. The irradiation at similar to 1000 degrees C did not induce hardening of stress-relieved (SR) W materials, but SR W materials tended to exhibit a decrease in the ultimate tensile strength (UTS) and an increase in total elongation (TE). The softening due to the recovery and recrystallization of SR W materials and the hardening due to the formation of irradiation defect clusters were balanced during irradiation at similar to 1000 degrees C, and ductility was exhibited without an increase in strength. (C) 2020 Elsevier B.V. All rights reserved.

Morphological and nanomechanical alteration of tungsten in extreme environments, like those in edge localized modes in nuclear fusion environments, up to 46.3 GWm −2 heat fluxes were experimentally simulated using electrothermal plasma. Surface and subsurface damage to the tungsten is seen mainly in the form of pore formation, cracks, and resolidified melt instabilities. Mirco voids, rosette-type microfeatures, core-shell structure, particle enrichment, and submicron channels all manifest in the damaged subsurface. The formation of voids in the subsurface was determined to originate from the ductile fracture of hot tungsten by plastic flow but not developed to cracking. The voids were preferentially settled in grain boundaries, interfaces. The directionality of elongated voids and grains is biased to the heat flow vector or plasma pathway, which is the likely consequence of the thermally driven grain growth and sliding in the high-temperature conditions. The presence of a border between the transient layer and heat-affected zone is observed and attributed to plasma shock and thermal spallation of fractural tungsten at high temperature. Plasma peening-like hardening effects in tungsten were observed in the range of 22.7–46.3 GWm −2 but least in the case of the lowest heat flux, 12.5 GWm −2 .

Summary form only given. Electrothermal plasma sources operating in the confined controlled arc discharge regime generate plasmas which can be used for a wide variety of applications, including materials deposition, mass acceleration devices, and high heat flux material exposure. Such plasmas are produced by capillary arc discharges-discharging high voltage across an insulated sleeve. Energy radiated to the walls of the insulated sleeve is absorbed by the material surface, ablating the material and forming a dense vapor. This vapor ionizes and forms a plasma which acts as a blackbody radiation source, which in turn can provide high heat flux to exposed surfaces. The generated plasma then exits the sleeve at high velocity. A substrate placed at the exit of the sleeve can act as a deposition surface for the ablated material. The electrothermal plasma source PIPE produces high density plasmas with heat fluxes on the order of gigawatts per square meter over periods of hundreds of microseconds. Sleeves of different materials can be machined and outfitted to the device, changing the composition of the plasma produced by their ablation. In addition, these sleeves can provide material surfaces for investigation of high heat flux plasma exposure. Thermal conductivity, ionization energy, and heat of vaporization are important material qualities of these sleeves when predicting the characteristics of produced plasmas. The high heat flux plasmas produced in the PIPE device have been exposed to various materials, including polycarbonate Lexan and elemental tungsten. The plasma exiting the PIPE device has been characterized by a range of optical techniques and compared to predicted characteristics. Electron temperature estimations and density measurements have been investigated with spectroscopic techniques. Spectral analysis has been performed to investigate dissociation of the plasma material. Velocity measurements of ejected plasma have been taken with high speed cameras. Material which is ablated or otherwise eroded is ejected from the device and collected on a substrate outside the device exit. Total ejected material has been determined by weight comparison of pre-and post-discharge insulating sleeves. Characteristics of the ejected material have been investigated with SEM and EDS techniques. Homogeneity and surface qualities of the redeposited material were investigated.

Summary form only given. Pulsed electrothermal (ET) plasma sources were originally used as ignition sources for electrothermal chemical weapons and secondary rail gun armature. An ET system consists of a discharge capacitor and accompanying electronics. The capacitor discharges through an electrode, normally comprised of tungsten, which sends current to ground through an insulating liner. The liner can be any insulating material that is subject to dissociation and ablation due to joule heating.

Electrothermal plasma sources operating in the confined controlled arc discharge regime produce heat fluxes in the range expected for hard disruptions in future large tokamaks. The radiative heat flux produced inside of the capillary discharge channel is from the formed high density (1023–1027/m3) plasma with heat fluxes of up to 125 GW/m2 over a period of 100 μs, making such sources excellent simulators for ablation studies of plasma-facing materials in tokamaks during hard disruptions. Graphite, beryllium, lithium, stainless steel, tungsten, copper, and molybdenum are among the materials proposed for use in fusion reactors. Computational experiments with the ETFLOW code using heat fluxes between 10 and 125 GW/m2 have shown low total erosion for the low-z materials Li, Be and C and higher erosion for high-z materials Fe, Cu, Mo and W. The time rate of material erosion for various ranges of heat fluxes shows increased erosion with time evolution over the 150 μs pulse length of the simulated disruption event. At the highest values of simulated heat flux, low-z materials were found to ablate almost identically. At all simulated values of heat flux, the ablation of high-z materials correlated positively with the z-number.

A new high energy density plasma deposition technique has been proposed and tested in order to investigate the formation of metal-vapor plasmas from a single or a mixed form of materials. The concept of electrothermal segmented plasma source (ETSPS) has been used for studies related to surface coatings, hardening, surface modification, ion implantation, materials synthesis, and the physics of complex mixed multi-component plasmas. The segmented source is a capillary discharge where the ablation liner is made from segments of either from same material or different material in each segment. Preliminary study to examine this technique for single and mixed material coatings has been successfully conducted. Surfaces of deposited substrates have been investigated and analyzed using high resolution optical microscopy, SEM, EDS and FIB milling near the edge. Experiments with Mo, Cu, Mo-Cu, and C-Mo-Cu segments have shown evidence of deposition on the titanium substrates.

Advanced fusion energy systems require the development of materials able to resist high operating temperatures, high neutron irradiation, and high thermo-mechanical stresses. Additional requirements for fusion materials include low absorption cross-sections, high thermal conductivity, and minimal swelling. Graphite, beryllium, lithium, stainless steel, tungsten, copper, and molybdenum are among the materials and alloy/ceramic components proposed for use in fusion reactors. Even with advances in the reliability of non-disruptive plasma systems, plasma-facing components, such as the diverter and first wall, must withstand abnormal hard disruption events as well as normal operational conditions. Sustained operation of fusion devices necessitates understanding long-term effects of materials response to high heat fluxes experienced during disruptions. The high heat flux effects on fusion reactor suitable materials is investigated herein by simulating heat flux deposition to materials of the interior liners inside electrothermal (ET) plasma sources, which produce high-density (10 23 -10 27 /m 3 ) plasmas with heat fluxes of up to 125 GW/m 2 over a period of 100μs relevant to expected heat fluxes during hard disruptions. Radiative heat flux to the inner wall of the ET source ablates the wall material, forming a dense vapor of excited atoms or molecules dissociated from the wall, followed by their ionization. The ETFLOW is a 1D time dependent code which models ET sources, the plasma formation and flow inside the source. Additionally, it calculates the incident and deposited heat flux, the amount of ablated mass, the pressure, velocity and number densities of the species. Erosion of selected materials has been studied over a range of heat fluxes comparable to expected heat fluxes during hard disruptions in future tokamaks. Computational experiments using heat fluxes between 10GW/m 2 and 125 GW/m 2 have shown low total erosion for low-z materials (Li, Be, C) and higher erosion for high-z materials (Fe, Cu, Mo, W). The time rate of material erosion for various ranges of heat fluxes shows increased erosion with time evolution over the 150 ms pulse length of the simulated disruption event. At the highest values of heat flux simulated, low-z materials were found to ablate almost identically. At all simulated values of heat flux, the ablation of high-z materials correlated positively with z-number.

In ablation controlled capillary discharges, the inner ablating sleeve is usually made of an insulator such as polyethylene or Lexan. There has been almost no use of metals, alloys, or semi-conductor materials in capillary discharges as most of the studies were devoted to the generation of high enthalpy plasma flows using polyethylene as the liner material. The work of Shcolnikov et al. [1995] provided a theoretical study in which the authors proposed using the ET plasma to accelerate powder particles in order to coat substrates and produce coatings with strong adhesion and low porosity.