Environmental barrier coatings for Zr-based materials are currently under development to reduce oxidation and embrittlement in light-water reactors. Chromium nitride is one such candidate for this application, particularly as accident-tolerant fuel cladding. However, quantifying the impact of coatings on the irradiation-induced creep of zircaloy (Zry) is critical as this mechanism often exceeds thermal creep rates under light-water reactor operating conditions and can be a limiting design characteristic. Additionally, examining irradiation effects in the microstructure at the coating interface is key to understanding the compatibility of the material system. To accelerate the experimental measurement of irradiation creep and microstructure evolution in CrN-Zry, compact, pressurized creep tubes were fabricated and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Miniature, thin-walled rodlets fabricated from annealed Zr-Sn barstock were coated with CrN using physical vapor deposition (PVD) to nominal thicknesses of 4 and 8 μm. Coated and uncoated rodlet specimens were internally pressurized and welded, generating nominal circumferential hoop stresses of 0, 90, or 180 MPa under 300∘C irradiation conditions. Twelve specimens were measured diametrically prior to irradiation using a low-cost, automated, contactless laser profilometer developed for this work. Specimens were irradiated in sealed capsules for one 25-day HFIR cycle, accumulating approximately 1.8×1021n/cm2 fast fluence (En>1.0MeV). The irradiated samples were retrieved and remeasured using the same profilometry system in a shielded hot cell facility. Irradiation creep between specimens was compared using standard statistical tests and showed that both thicknesses of CrN coating had a negligible effect on the irradiation creep strain of the Zry material. Microstructure characterization of pre- and post-irradiated CrN-Zry specimens showed minimal changes due to irradiation but did show a substantial O-rich region at the Zry-CrN interface.

Tube processing and aging effects in FeCrAl ODS steels are investigated in four mechanical alloyed ferritic ODS steels, Fe15Cr (SP2), Fe15Cr5Al (SP4), Fe15Cr7Al (SP7) and Fe18Cr7Al (SP11). These steels were made into 0.3mm thick plates by simulated tube processing (STP). Strengthening after partial recrystallization was achieved after the last cold rolling and heat treatment step. However, the ductility reduced about one third of the as-extruded steels. The STPed steels were aged at 475{\deg}C in sealed vacuum tubes up to 2000 hrs and 10000 hrs, respectively. The yield stress and elongation were investigated by tensile tests. The results revealed that all the STPed steels fractured in a ductile manner irrespective of aging conditions. Aging hardening and ductility reduction in STPed steels are similar to as-extruded ones. The STPed ODS steels showed similar ageing embrittlement resistance as as-extruded steels, but much higher than the non-ODS steels. The aging hardening based on cut-through and bow-pass mechanisms were discussed. The time dependent hardening of overaged steel (beta prime only) was analyzed as well.

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.

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.

Focused ion beam surface preparation for plasma facing materials Tungsten is one of the most important nuclear fusion structural materials for use as a plasma facing component in fusion reactors. However, in a critical range of temperatures a unique surface morphology known as “nanofuzz” forms when exposed to He-containing plasma. The so-called nanofuzz grows with increased exposure times, yet the tendrils remain in the size range of 10s of nanometers of diameter, forming a many-microns thick layer [1]. Understanding the exact formation mechanism of nanofuzz requires identifying structures and surface morphology within the top 10 nanometers of a tungsten sample after exposure to various He plasma and temperature conditions.

A mechanically alloyed austenitic stainless steel (MA304LZ) was produced from pre-alloyed SUS304L powder with a small amount of Zr addition. The yield stress of MA304LZ was more than 3 times larger than that of SUS304L or 316L, while total elongation was reduced to about one third of the conventional steels. Microstructure analysis revealed an average grain size of 0.42 mu m in MA304LZ and about 34/30 mu m in SUS304L/316 L. In MA304LZ, two types of precipitates were observed; inhomogeneously distributed fine precipitates with an average size of 6.0 nm and homogeneously distributed coarse precipitates (d > 20 nm) with an average size of 47 nm. The strengthening mechanism of MA304LZ was discussed on the bases of Hall-Petch and Orowan equations, and the strengthening of MA304LZ was attributed mostly to refined grains. The dislocation barrier strength factor, alpha, is estimated to be 0.277 for the Zr-rich precipitates in MA304LZ.

A mechanically alloyed austenitic stainless steel (MA304LZ) was produced from pre-alloyed SUS304L powder with 0.7% wt of Zr addition.The precipitates had a bimodal size distribution causing strengthening of MA304LZ with 3.3 times larger yield stress than SUS304L, although the contribution of fine grains of MA304LZ to the strengthening was larger than the dispersion of precipitates.For precipitate characterization, two types of extraction replica samples were produced; 1) Thin carbon film and 2) Electrochemically dissolved residue.Extraction replica residue was analyzed using XRD and both types were attached to copper grids for TEM, EDS, and EELS analysis.The previously reported Zr-rich precipitates were revealed to be ZrO 2 (zirconia) with no evidence of ZrN or ZrC.

A mechanically alloyed austenitic stainless steel (MA304LZ) was produced from pre-alloyed SUS304L powder with a small amount of zirconium addition. Coupon-type specimens of MA304LZ and SUS304L steels were subjected to hot water at 300 ℃/25 MPa and supercritical pressurized water (SCW) at 500 ℃/25 MPa for 1000 hr. MA304LZ is significantly less susceptible to corrosion weight gain in SCW than SUS304L which follows the parabolic rule between weight gain and elapsing time. The reduction of weight gain in MA304LZ can be attributed to much smaller grains which enhance chromium diffusion through grain boundaries and consequently accelerate the formation of a protective chromium oxide layer.

The nonmagnetic austenite steels are the jacket materials for low-temperature superconductors of fusion reactors. The present work provides evidences that austenites transform to magnetic martensite when deformation with a high-strain is imposed at 77 K and 4.2 K. The 4.2 K test is characterized by serrated yielding that is related to the specific motion of dislocations and phase transformations. The in-situ transmission electron microscope (TEM) observations in nanoscale reveal that austenites achieve deformation by twinning under low-strain conditions at deep cryogenic temperatures. The generations of twins, martensitic transformations, and serrated yielding are in order of increasing difficulty. (C) 2018 Elsevier B.V. All rights reserved.

In order to investigate the effects of oxide particles on radiation response such as hardness change and microstructural evolution, three types of oxide dispersion strengthened (ODS) ferritic steels (named Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS), mostly strengthened by Y-Ti-O, Y-AI-O and Y-Zr-O dispersoids, respectively, were simultaneously irradiated with iron and helium ions at 550 degrees C up to a damage of 30 dpa and a corresponding helium (He) concentration of similar to 3500 appm to a depth of 1000-1300 nm. A single iron ion beam irradiation was also performed for reference. Transmission electron microscopy revealed that after the dual ion irradiation helium bubbles of 2.8, 6.6 and 4.5 nm in mean diameter with the corresponding number densities of 1.1 x 10(23), 2.7 x 10(22) and 3.6 x 10(22) m(-3) were observed in Y-TiODS, Y-Al-ODS and Y-Al-Zr-ODS, respectively, while no such bubbles were observed after single ion irradiation. About 80% of intragranular He bubbles were adjacent to oxide particles in the ODS ferritic steels. Although the high number density He bubbles were observed in the ODS steels, the void swelling in Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS was still small and estimated to be 0.13%, 0.53% and 0.20%, respectively. The excellent swelling resistance is dominantly attributed to the high sink strength of oxide particles that depends on the morphology of particle dispersion rather than the crystal structure of the particles. In contrast, no dislocation loops were produced in any of the irradiated steels. Nanoindentation measurements showed that no irradiation hardening but softening was found in the ODS ferritic steels, which was probably due to irradiation induced dislocation recovery. The helium bubbles in high number density never contributed to the irradiation hardening of the ODS steels at these irradiation conditions. (C) 2018 Elsevier B.V. All rights reserved.