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AbstractThe United States and Japan have collaborated on fusion materials research in a series of agreements reaching back to 1981. The PHENIX collaboration is the latest U.S.-Japan project which spans 2013 to 2019 and has the goal of assessing technical feasibility of tungsten-based, helium-cooled plasma-facing component concepts for a demonstration fusion power reactor (DEMO). Task 2 within the PHENIX project is focused on evaluating the neutron irradiation effects in tungsten. For tungsten, the transmutation to Re and Os is at least as important to determining its properties after irradiation as the displacement damage, and the transmutation rate depends on the energy spectrum of the reactor. A large-scale, instrumented irradiation capsule with thermal neutron shielding to better mimic fusion conditions was irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. The tungsten specimens were irradiated in different temperature zones between 500°C and 1200°C to doses of ~0.2 to 0.7 displacements per atom. More than 20 varieties of pure tungsten and tungsten alloys were included in the irradiation, and they were evaluated in the 3025E hot-cell facility and at the Low Activation Materials Development and Analysis Laboratory. The elevated temperature tensile, fracture toughness, hardness, thermal conductivity, electrical resistivity, density, elemental composition, and microstructure properties of the irradiated materials are being collected. This paper overviews the experimental design, specimen matrix, and the initial results of postirradiation examinations.

AbstractThe annealing effects on deuterium (D) retention for 0.1–1.0 dpa iron (Fe) ion damaged W were studied as a function of annealing duration. The D2spectra for Fe damaged W with lower defect concentration showed that D trapped by vacancy clusters was clearly decreased as increasing annealing duration due to the recovery of vacancy clusters. On the other hand, at higher defect concentration, the desorption peak of D trapped by voids was shifted toward higher temperature side, which would be caused by aggregation of vacancies and vacancy clusters. It can be said that the recovery and aggregation behavior of defects are controlled by defect concentration. By disappearing of desorption of D trapped by vacancy clusters after annealing for longer duration, the desorption of D trapped by vacancies was increased, which could be explained by following two possibilities. One is that the retention of hydrogen isotope trapped by monovacancy was increased. The other is that number of vacancies during annihilation process of vacancy cluster were formed by annealing.

AbstractThe PHENIX Project is a 6-year U.S./Japan bilateral, multi-institutional collaboration program for the Technological Assessment of Plasma Facing Components for DEMO Reactors. The goal is to address the technical feasibility of helium-cooled divertor concepts using tungsten as the armor material in fusion power reactors. The project specifically attempts to (1) improve heat transfer modeling for helium-cooled divertor systems through experiments including steady-state and pulsed high-heat-load testing, (2) understand the thermomechanical properties of tungsten metals and alloys under divertor-relevant neutron irradiation conditions, and (3) determine the behavior of tritium in tungsten materials through high-flux plasma exposure experiments. The High Flux Isotope Reactor and the Plasma Arc Lamp facility at Oak Ridge National Laboratory, the Tritium Plasma Experiment facility at Idaho National Laboratory, and the helium loop at Georgia Institute of Technology are utilized for evaluation of the response to high heat loads and tritium interactions of irradiated and unirradiated materials and components. This paper provides an overview of the progress achieved during the first 3 years and discusses the plan for the remainder of the project.

As a preliminary investigation of process development for SiC fiber-reinforced SiC matrix composites (SiCf/SiC composites) via a pressureless liquid phase sintering (PLPS) method, the manufacturing process parameters of monolithic SiC fabricated with/without very low pressure during the sintering under high temperature (1850 and 1900 °C) were optimized. Two kinds of sintering additive system (only Al2O3and Al2O3–Y2O3mixtures) can be achieved full densification corresponding to 99% of the theoretical density without high pressure during the sintering. Monolithic SiCs of both sintering additive systems applied with prepressure of 20 or 40 MPa possessed the flexural strength and porosity comparable to those of conventional liquid-phase sintered SiC with high pressure during the sintering. The low pressure (0.63 MPa) during sintering reduced the densification in particular for the materials sintered with just Al2O3. The additions of BN particles led to a gradual decrease in mechanical strength with an increase in porosity.

Irradiation damage produced by neutrons or energetic particles lead to changes of physical- and mechanical-properties of SiC. Radiation hardening and fracture toughness changing of SiC were clarified by indentation method previously. However, the mechanism studies have received little alteration. The purpose of this study is to improve the understanding of the mechanisms of mechanical property changes under irradiation. In this paper, the microstructural observation beneath and near an indentation will be used to infer mechanisms of radiation hardening and toughening. Indenting polycrystalline SiC creates deformation and cracking in the plastically deformed region. In the case of irradiated SiC, however, small-sized deformation zone was observed below contact indent, which resulted in the restricted size of residual impression. Additionally, the indentation cracks showed complex propagation behaviors such as deflecting, branching and microcracking.

SiC-composites are candidate structural materials for high temperature applications such as gas turbines. For this purpose, a suitable coating against high temperature oxidation is essential. The coating material selection was made by phase stability/durability under service environment, CTE matching, low elastic modulus and chemical compatibility with SiC. This work is to evaluate the microstructural quality of mullite and erbium silicate coatings on SiC fiber bonded composites. In this work, SiC-fiber bonded 2D-composite (TyrannohexTM) by Ube Industries Ltd., were coated with mullite and erbium silicate by plasma spraying. The thickness of mullite and erbium silicate coatings was determined to be 120 microns and 220 microns, respectively. The fabricated samples were diamond polished for cross-sectional analyses by optical and scanning electron microscopy. The coating-substrate interfaces have been found bit undulating. The bonding between mullite and the substrate was found insufficient as compared to that observed in the erbium silicate coated substrate. The cross-sectional analyses of erbium silicate coating revealed the presence of throughthickness micro- and macro-cracks. The width of the macrocracks was found in the range of 1-3 microns. It is conceived that the evolution of large concentration of through-thickness-cracks is due to the stresses in the coating. The most probable sources of stresses are the precipitation of various second phases in the coating materials and the thermal expansion mismatches that arise during cooling. The mullite coating on the other hand is not found with macrocracks and on the whole the cohesion within the mullite coating is rather strong. The cross-sectional examination also revealed the irregularly distributed pores and cavities in both coating systems that are typical of thermal sprayed coatings. The surfaces of the coatings are characterized of typical lamellar microstructure formed by the flattened platelets of individual particles. The size of these droplets ranges from submicrometer to several micrometers.

AbstractReduced activation ferritic/martensitic steels (RAFs) are leading candidates for blanket and first wall of fusion reactors where effects of displacement damage and helium production are important subjects to be investigated. To obtain systematic and accurate information of microstructural response under fusion environment, dual-ion irradiation method was applied. In order to estimate the microstructural response under fusion neutron irradiation environment, ion-beam irradiation was carried out with helium and metallic self ions. The study is focused on JLF-1 single- and dial-ion irradiated up to 60 dpa at 693, 743 and 793 K. The damage rate and helium injection rate were 1.0 × 10–3dpa/sec and 15 × 10–3appm He/sec. At 743 K, void cavity structure was observed under dual-ion irradiation where the contribution of void structure on hardening was not so significant. Irradiation hardening and swelling were depended for the case of dual-ion irradiation. It is attempted to quantitatively relate the dislocation and cavities to the irradiation induced hardening.

AbstractThe microstructural evolution of SiC/SiC composites after Si2+with/without He+ion irradiation was studied using transmission electron microscopy. The temperature, displacement damage level, and He/dpa ratio were 1273/1673K, 10/100dpa and 0/60appmHe/dpa, respectively. In 10dpa single-ion irradiation, no cavity was detected at 1273 and 1673K. But cavities were observed locally at 1673K, 100dpa. In dual-ion irradiation, cavities were observed at 1673K, 100dpa. Helium bubbles (d<5nm) were formed densely on {111} faulted planes in the fiber and matrix. And lens-shaped cavities (major axis 2a=20-50nm) were formed on grain boundaries in the matrix. The swelling by cavities in CVI matrix is about 0.5% at 80dpa and 0.7% at 130dpa. Loss of PyC layer beneath the irradiated surface was observed (single-ion: about 500nm, dual-ion: about 1 μm). And the thickness of the PyC layer expands after single/dual-ion irradiation (single-ion: 12%, dual-ion: 29% increase). But Tyranno-SA/PyC/CVI composites shows showed better microstructural stability than expected at 1673K.

Three kinds of atmospheres (air, highly-pure Ar and ultra highly-pure Ar gas) with different oxygen partial pressures were applied to investigate the tensile properties and creep behavior of SiC fibers such as Hi-NicalonTM and TyrannoTM-SA. These fibers were annealed and crept at elevated temperatures ranging from1273-1773 K in such environments. After annealing at 1773 K, the room temperature tensile strengths of SiC-based fibers decreased with decreasing the oxygen partial pressure and the near stoichiometric fiber TyrannoTM-SA shows excellent strength retention. At temperatures above the 1573 K, the creep resistance of SiC fibers evaluated by bending stress relaxation (BSR) method under high oxygen partial pressure was lower than that of in low oxygen partial pressure. The microstructural features on these fibers were examined by scanning electron microscopy (SEM) and X-ray diffraction (XRD).

AbstractMechanical properties of silicon carbide composites reinforced with highly crystalline fibers and fabricated by the chemical vapor infiltration method were evaluated. Materials used were SiC/SiC composites reinforced with unidirectional Hi-Nicalon Type-S fibers and unidirectional Tyranno SA fibers with various fiber/matrix interphase. Also, SiC/SiC composites reinforced with plain weave Tyranno SA fibers with carbon or multilayers of silicon carbide and carbon interphase were evaluated. In-plane tensile, transthickness tensile and interlaminar shear properties were evaluated by the in-plane tensile test, the transthickness tensile test, the diametral compression test and the compression test of double-notched specimens.The elastic modulus and proportional limit stress were improved by using high purity silicon carbide fibers. The in-plane tensile properties were insensitive to carbon interphase thickness for a range of thicknesses between 30 and 230 nm. It was found that the in-plane tensile strength of composites containing multilayers of silicon carbide and carbon coating of fibers and fiber bundles was superior to that of composites with carbon alone. Transthickness tensile strength and shear strength of high purity silicon carbide composites were successfully evaluated.

AbstractSilicon carbide (SiC) fiber-reinforced SiC-matrix ceramic composite (SiC/SiC composite) is an attractive material for blanket/first wall structures in fusion power devices. Recent extensive materials R&D efforts are producing advanced SiC/SiC composites substantially different from conventional materials in terms of baseline properties as well as irradiation stability. This paper provides a summary of the recent fusion-relevant progress in development and irradiation effect studies of SiC/SiC composites achieved through Japanese programs and Japan/US collaborative JUPITER(-II) program for fusion materials and blanket engineering.

SiC fiber reinforced-SiC ceramic matrix (SiCf/SiC) composites are one of the candidate structural materials for nuclear fusion reactors systems because of their low activation, mechanical property and chemical stability at high temperature. High thermal conductivity of SiCf/SiC composites is required to reduce heat load and to provide high thermal conversion efficiency. The objective of this study is to determine thermal conductivity on grain size dependence of SiC ceramics which is the matrix in composites. b-SiC powder was sintered by hot-press method. Al2O3 and Y2O3 were used as sintering additives. The grains size was grew with increase of holding time at 1900 degC. Thermal conductivity of SiC ceramics was gradually increased up to 112 W/mK. Thermal conductivity of SiC ceramics is dependent on growth of grain size due to the scattering of the phonon by the grain boundaries leads to a size-dependence.

Silicon carbide composites (SiC/SiC) are attractive candidate materials for structural and functional components in fusion energy systems. The effect of neutron irradiation on damage tolerance of the nuclear grade SiC/SiC composites (plain woven Hi-Nicalon Type-S reinforced CVI matrix composites multilayer interphase and unidirectional Tyranno-SA3 reinforced NITE matrix with carbon mono-layer interphase) was evaluated by means of miniaturized single-edged notched beam test. No significant changes in crack extension behavior and in the load-loadpoint displacement characteristics such as the peak load and hysteresis loop width were observed after irradiation to 5.9 x 1025 n/m2 (E > 0.1 MeV) at 800degC and to 5.8 x 1025 n/m2 at 1300degC. By applying a global energy balance analysis based on non-linear fracture mechanics, the energy release rate for these composite materials was found to be unchanged by irradiation with a value of 3+-2 kJ/m2. This has led to the conclusion that, for these fairly aggressive irradiation conditions, the effect of neutron irradiation on the fracture resistance of these composites appears insignificant.