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Unidirectional SiCf/SiC-matrix composites with a pyrocarbon interface were prepared by hot-pressing SiC nano-powder via a liquid phase of sintering additives under high temperature (1900 °C) and applied pressure (20 MPa). The SiC reinforcing fiber volume fraction of the composites varied from 0 to 78 vol% and was found to have significant effects on densification and mechanical properties, where changes in failure mode, proportional limit stress, ultimate strength, and elastic modulus were observed. The mechanical properties were significantly enhanced with an increase in the fiber volume fraction until a moderately high fiber content was obtained, after which the ultimate strength of the composite degraded. These observations were related to the open porosity within the fiber bundles. Further, fiber strength was affected by fiber damage caused by contact with the surrounding matrix due to poor densification of the SiC nano-powder.

Replacement of conventional Zircaloy fuel cladding with silicon carbide (SiC) fuel cladding is expected to significantly decrease the amount of hydrogen generated from fuel claddings by the reaction with steam during severe accidents. One of their critical issues addressed regarding practical application has been hydrothermal corrosion. Thus, the corrosion resistant coating technology using a Ti-Cr multilayer was developed to suppress silica dissolution from SiC fuel cladding into reactor coolant under normal operation. The effect of radiation on adhesion of the coating to SiC substrate and its microstructure characteristics were investigated following Si ion irradiation at 573 K up to 3 dpa for SiC. Measurement of swelling in pure Ti, pure Cr and SiC revealed that the maximum inner stress attributed to the swelling difference was generated between the coating and SiC substrate by irradiation of 1 dpa. No delamination and cracking were observed in cross-sectional specimens of the coated SiC irradiated up to 3 dpa. According to analyses using transmission electron microscopy, large void formation and cascade mixing due to irradiation were not observed in the coating. The swelling in the coating at 573 K was presumed to be caused by another mechanism during radiation such as point defects rather than void formation.

A tungsten (W) material is a candidate for the first wall and silicon carbide (SiC) composites are candidates for the structural materials applied in nuclear fusion. SiC fiber-reinforced W composites are also developed for nuclear fusion applications. An effective diffusion barrier is required to prevent reaction between W and SiC. Therefore, in this work, advanced ceramics coatings, such as oxides (ZrO2, TiO2 and Er2O3), nitrides (ZrN and TiN), carbides (TiC and ZrC) were chosen to assess abilities to suppress the reactions. Various films were coated on a CVD (chemical vapor deposition)-SiC plate using the dipping method. Additionally, nitrides coatings prepared by the sputtering method were also investigated in this work. Then evaluations were carried out by joining the coated CVD-SiC plates with W foils. Only the multi-dipped Er2O3 coating and the sputtered nitrides worked well compared with the other coatings. For the other oxide coatings, reactions were identified between oxides and SiC, and for the dipped nitrides and carbides films, cracks were observed on the coating, generated from the coefficient of thermal expansion (CTE) mismatch with the SiC substrate and volume change for the oxides changing to nitrides and carbides. This work provides suggestions about choosing an appropriate interface material between SiC and W.

This study fills a knowledge gap regarding neutron-irradiation resistance of SiC joints for nuclear applications, by investigating high-dose neutron irradiation effects on the strength of selected joints and low-dose neutron irradiation effects on recently developed joints fabricated by state of the art processing methods. The joining methods used for the high-dose radiation study included pressure-assisted liquid-phase sintering (LPS) of SiC nanopowder, pressureless calcia-alumina glass ceramics joining, and reaction sintering of Ti-Si-C powders with hot-pressing. The joints were neutron-irradiated at 530 °C to 20 displacements per atom (dpa). Other joining methods included low-pressure LPS of cold-pressed SiC green body, pressureless reaction sintered Ti-Si-C powder joint, spark plasma–sintered Ti diffusion bond, and hot-pressed Ti diffusion bond, which were irradiated at ∼500 °C to ∼2 dpa. There was no notable degradation of torsional strengths of the joints following the high-dose irradiation. The irradiation-induced degradation at low neutron dose was highly dependent on joint type.

The 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 s...

Various oxides and SiC have been proposed as candidate materials for electrical insulation coatings and inserts to suppress the magnetohydrodynamic (MHD) pressure drop in liquid metal fusion blanket systems. From the early stage of the MHD insulator development, the magnitudes of the radiation-induced conductivities (RICs), i.e., increases in conductivities by radiation excitation, have been evaluated on bulk and coated specimens by using several radiation sources. While the insulating performances of the oxide coated and SiC plate specimens are inferior to those of high quality and purity bulk materials due to cracks, pits, sintering additives, etc., the results indicate that the RIC phenomenon would not prevent the MHD insulators from achieving the required performances (10−3–10−2 S/m for coatings, 102 S/m for inserts) at the first wall of the blanket (500–700 °C, several kGy/s). The recent design investigation of blanket modules for DEMO reactor conditions provides more precise information for prediction of the performances in reactors and conditions to be simulated in the future irradiation damage studies. In the case of the coatings, the suppression of recoiled Li from the liquid metal breeder/coolant might be an issue to be considered in the development.

Silicon carbide composites are expected for light water reactors. The objective is to understand the steam oxidation behavior and the high-temperature water corrosion behavior of the liquid phase sintering silicon carbide and to develop the liquid phase sintering silicon carbide composites, which are stable at the high-temperature water conditions in normal operation and the high-temperature steam conditions in a severe accident. The steam oxidation experiments were carried out at 1200 and 1400 °C. The high-temperature water corrosion experiments were carried out at 320 and 360 °C. The formation of the silicate, which is expected to have excellent resistance to the steam, was confirmed following the steam exposure at 1400 °C. High-temperature water corrosion resistance was improved by the formation of Yttrium Aluminum Garnet at the grain boundary. The particle-dispersion silicon carbide composite tubes with the modified condition were developed, and the thermal shock experiments from 1200 °C to ambient temperature were carried out. The composite tubes showed excellent oxidation and thermal shock resistance. The particle-dispersion liquid phase sintering silicon carbide composites with the modified condition are promising materials for light water reactors.

To study the effect of the content of chromium (Cr) in the tungsten (W) matrix on the vacancy formation and retention of hydrogen isotopes, the samples of the W-0.3 at.% Cr alloy were irradiated with 6.4 MeV Fe ions in the temperature range of 523–1273 K to a damage level of 0.26 displacement per atom (dpa). These displacement-damaged samples were exposed to D2 gas at a temperature of 673 K and a pressure of 100 kPa to decorate ion-induced defects with deuterium. The addition of 0.3 at.% Cr into the W matrix resulted in a significant decrease in the retention of deuterium compared to pure W after irradiation especially at high temperature (≥773 K). Positron lifetime in W-0.3 at.% Cr alloy irradiated at 1073 K was almost similar to that for non-irradiated one. These facts indicate the suppression of the formation of vacancy-type defects (monovacancies and vacancy clusters) by 0.3 at.% Cr addition, which leads to the significant reduction in deuterium retention in W-0.3 at.% Cr alloy.

Silicon carbide (SiC) fiber reinforced SiC matrix composites continue to undergo development for fusion applications worldwide because of inherent advantages of the material including low activation, high temperature capability, relatively low neutron absorption, and radiation resistance. This paper presents an international overview of recent achievements in SiC-based composites for fusion applications. Key subjects include applications in fusion reactors, high-dose radiation effects, transmutation effects, material lifetime assessment, and development of joining technology (processing, test method development, irradiation resistance, and modeling capability). This paper also discusses synergy among research for fusion materials and non-fusion materials (for fission and aerospace applications). Finally, future research directions and opportunities are proposed.

The R&D on SiC/SiC composites under the broader approach (BA) activities between Japan and the EU for fusion DEMO developed a fundamental database of mechanical (Task-1) and physical/chemical (Task-2) properties, with a primary target of the application of SiC/SiC composites as functional structure to be used in the dual coolant breeding blanket concept. This paper aims to summarize previous 10-years activities of the R&D of Japan and to provide the key deliverables toward the DEMO design. In Task-1, good creep and fatigue durability were first demonstrated. Besides, in-plane and inter-laminar strength anisotropy maps at elevated temperatures were comprehensively identified. In parallel, the irradiation effects of SiC materials were specifically determined as input parameters of the analytical model to provide for the irradiation-induced residual stresses. In Task-2, the apparent dose-dependence of the radiation-induced electrical conductivity and the indicative radiation-induced electrical degradation was identified by various irradiation sources. In addition, good gas confinement was identified. Furthermore, no accelerated corrosion for duration of 3000 h at below 1173 K was first demonstrated. With these achievements, it is suggested that the in-vessel component technology, e.g., material corrosion database development, activated corrosion product evaluation code development, compact module tests for validation of the key functions of the components, technology integration assessment for fusion nuclear tests, etc., should be further developed toward DEMO in near-term.

Silicon carbide (SiC) porous-matrix composites lacking a fiber/matrix interface and incorporating continuous unidirectional SiC fibers as reinforcements were developed using sandwiched layers with non-infiltration fiber bundles and a porous matrix in a laminate to achieve crack deflection. Excellent control of the porosity was achieved by varying the amount of carbon powder added. The matrix porosity was characterized in terms of its microstructure and mechanical properties. The SiC porous-matrix composite with an open porosity of 22%, which was formed with 40 vol% carbon powder, had a flexural strength of 253 ± 31 MPa and exhibited non-brittle fracture behavior up to a stress maximum, followed by a gradual depletion of the fracture energy, decreasing the stress on continued deformation. Significant changes in the mechanical properties, such as the flexural strength and fracture energy, of the fabricated SiC porous-matrix composite were not observed following exposure to air at high temperature (up to 1100 °C).

Silicon carbide (SiC) and its composites are candidate materials for the blanket components and for the first wall in a fusion reactor. If the SiC is used without any armor materials for the first wall, it is exposed by helium (He) plasma as well as hydrogen plasma. Characteristic surface morphology changes are reported for various materials by He plasma exposure. Thus, we exposed SiC specimens to He or simultaneous deuterium (D) and He (D + He) plasma by various conditions and then observed surface morphology changes by SEM. As a result, needle-like structures and whiskers-like structures at the tip were formed in He plasma and D + He irradiation, while only needle-like structures were formed in D plasma. Therefore, it indicated that the effects of He were attributed to form whiskers-like structures. Although the structures are different among He plasma, simultaneous D + He plasma and D plasma irradiations, sputtering is considered to be a dominant process for the formation of the structure formation. However, the effects of He atoms in the structure could also be attributed to form whiskers-like structures. Keywords: Helium induced nanostructure, Morphology change, Silicon carbide

To evaluate Helium (He) effect on hydrogen isotope retention in tungsten (W), He+ was introduced into W bulk by 201 – 1000 keV He+, or W surface by 3 keV He+ for 6 MeV Fe ion damaged W at room temperature. The deuterium (D) retention behavior was evaluated by thermal desorption spectroscopy (TDS). In addition, the amount of tritium (T) at surface and bulk were separately evaluated by beta-ray induced X-ray spectroscopy (BIXS). The experimental results indicated that the formation of He-void complexes reduced the D trapping in vacancies and voids which have higher trapping energy by the bulk He retention. The BIXS measurement also supported the He enhanced the D reduction in the W bulk region. On the other hand, the He ion irradiation near the surface region enhanced D trapping by dislocation loops or surface, indicating the existence of He near surface interfered the D diffusion toward the bulk. It was concluded that the He existence in bulk or surface will significantly change the D trapping and diffusion behavior in damaged W. Keywords: He in damaged W, Hydrogen isotope retention behavior, Damaged W, Fusion

As silicon carbide (SiC) is an inert material, it has attracted attention as an alternative material for the core components of nuclear fuel cladding. However, the corrosion of SiC under neutron irradiation has been reported, and this has been a bottleneck issue. Here, we semi-quantitatively introduced point defects into single crystal 3C-SiC layers by ion irradiation, and investigated their corrosion behavior in terms of the electrochemical activity associated with the point defects. The results showed a shift in corrosion potential and an increase in the corrosion current due to the introduction of defects. However, in the case of aluminum (Al)-doped SiC, these changes were more moderate than in the cases of the nitrogen (N)- or boron (B)-doped material, implying that Al-doped SiC has a better tolerance against corrosion. Keywords: SiC, Point defects, Corrosion, Dopant

The aim of this work was to investigate and discuss the microstructure and interface reaction of a calcia-alumina based glass-ceramic (CA) with SiC. CA has been used for several years as a glass-ceramic for pressure-less joining of SiC based components. In the present work, the crystalline phases in the CA glass-ceramic and at the CA/SiC interface were investigated and the absence of any detectable amorphous phase was assessed. In order to provide a better understanding of the effect of irradiation on the joining material and on the joints, Si ion irradiation was performed both on bulk CA and CA joined SiC. CA glass-ceramic and CA joined SiC were both irradiated with 5.1 MeV Si2+ ions to 3.3 × 1020 ions/m2 at temperatures of 400 and 800 °C at DuET facility, Kyoto University. This corresponds to a damage level of 5 dpa for SiC averaged over the damage range. This paper presents the results of a microstructural analysis of the irradiated samples as well as an evaluation of the dimensional stability of the CA glass-ceramic and its irradiation temperature and/or damage dependence.

The 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 D2 spectra for Fe damaged W with lower defect concentr...

The 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 respo...

Monolithic silicon carbide (SiC) to SiC plate joints were fabricated and irradiated with neutrons at 270–310 °C to 8.7 dpa for SiC. The joining methods included solid state diffusion bonding using titanium and molybdenum interlayers, SiC nanopowder sintering, reaction sintering with a Ti-Si-C system, and hybrid processing of polymer pyrolysis and chemical vapor infiltration (CVI). All the irradiated joints exhibited apparent shear strength of more than 84 MPa on average. Significant irradiation-induced cracking was found in the bonding layers of the Ti and Mo diffusion bonds and Ti-Si-C reaction sintered bond. The SiC-based bonding layers of the SiC nanopowder sintered and hybrid polymer pyrolysis and CVI joints all showed stable microstructure following the irradiation.

To investigate systematically the effects of Re and other elements on deuterium (D) retention, W and binary W-alloys (Re, Mo and Ta) were irradiated with 6.4 MeV Fe ions at high temperatures (1073–1273 K) and then exposed to D2 gas at 673 K. Depth profiles of D were measured by nuclear reaction analysis (NRA), and D retention was determined by thermal desorption spectrometry (TDS) and NRA. The addition of 5 at.% Re into W reduced the content of D trapped at radiation-induced defects created by irradiation with the Fe ions at 1273 K to the peak damage level of 5 displacements per atom (dpa). At Re fractions of 1, 3 and 5 at.%, comparable effects on D retention were observed after irradiation to 0.5 dpa at 1073 K. The addition of Mo and Ta to W resulted in no visible effects in D retention.

This paper reviews recent results pertaining to hydrogen isotope behavior in neutron and heavy ion damaged W. Accumulation of damage in W creates stable trapping sites for hydrogen isotopes, thereby changing the observed desorption behavior. In particular, the desorption temperature shifts higher as the defect concentration increases. In addition, the distribution of defects throughout the sample also changes the shape of TDS spectrum. Even if low energy traps were distributed in the bulk region, the D diffusion toward the surface requires additional time for trapping/detrapping during surface-to-bulk transport, contributing to a shift of desorption peaks toward higher temperatures. It can be said that both of distribution of damage (e.g. hydrogen isotope trapping sites) and their stabilities would have a large impact on desorption. In addition, transmutation effects should be also considered for an actual fusion environment. Experimental results show that production of Re by nuclear reaction of W with neutrons reduces the density of trapping sites, though no remarkable retention enhancement is observed.

Effects of annealing after/under iron (Fe) ion irradiation on deuterium (D) retention behavior in tungsten (W) were studied. The D2 TDS spectra as a function of heating temperature for 0.1 dpa damaged W showed that the D retention was clearly decreased as the annealing temperature was increased. In particular, the desorption of D trapped by voids was largely reduced by annealing at 1173 K. The TEM observation indicated that the size of dislocation loops was clearly grown, and its density was decreased by the annealing above 573 K. After annealing at 1173 K, almost all the dislocation loops were recovered. The results of positron annihilation spectroscopy suggested that the density of vacancy-type defects such as voids, was decreased as the annealing temperature was increased, while its size was increased, indicating that the D retention was reduced by the recovery of the voids. Furthermore, it was found that the desorption temperature of D trapped by the voids for damaged W above 0.3 dpa was shifted toward higher temperature side. These results lead to a conclusion that the D retention behavior is controlled by defect density. The D retention in the samples annealed during irradiation was less than that annealed after irradiation. This result shows that defects would be quickly annihilated before stabilization by annealing during irradiation.

Hydrothermal corrosion of four types of the silicon carbide (SiC) to SiC plate joints were investigated under pressurized water reactor and boiling water reactor relevant chemical conditions without irradiation. The joints were formed by metal diffusion bonding using molybdenum or titanium interlayer, reaction sintering using Ti–Si–C system, and SiC nanopowder sintering. Most of the joints withstood the corrosion tests for five weeks. The recession of the SiC substrates was limited. Based on the recession of the bonding layers, it was concluded that all the joints except for the molybdenum diffusion bond are promising under the reducing environments without radiation. The SiC nanopowder sintered joint was the most corrosion tolerant under the oxidizing environment among the four joints.

An enhancement of the dissolution of high-purity 3C-SiC in hot water (320 °C, 20 MPa: relevant to the light-water reactor coolant condition) is demonstrated after 5.1 MeV Si-ion irradiation. Optical spectrometry and Kelvin force microscopy revealed the creation of interband-defect localized states within the bandgap. The dissolution rate was found to be dependent on the irradiation fluence, irradiation-induced volume expansion, and the photoluminescence quenching. An annealing study showed prevention of irradiation-enhanced dissolution with the recovery of most defects. These results show that the dissolution rates of irradiated SiC are increased with the population of irradiation-induced defects.

Irradiation creep beyond the transient regime was investigated for various silicon carbide (SiC) materials. The materials examined included polycrystalline or monocrystalline high-purity SiC, nanopowder sintered SiC, highly crystalline and near-stoichiometric SiC fibers (including Hi-Nicalon Type S, Tyranno SA3, isotopically-controlled Sylramic and Sylramic-iBN fibers), and a Tyranno SA3 fiber–reinforced SiC matrix composite fabricated through a nano-infiltration transient eutectic phase process. Neutron irradiation experiments for bend stress relaxation tests were conducted at irradiation temperatures ranging from 430 to 1180 °C up to 30 dpa with initial bend stresses of up to ∼1 GPa for the fibers and ∼300 MPa for the other materials. Initial bend stress in the specimens continued to decrease from 1 to 30 dpa. Analysis revealed that (1) the stress exponent of irradiation creep above 1 dpa is approximately unity, (2) the stress normalized creep rate is ∼1 × 10 −7 [dpa −1 MPa −1 ] at 430–750 °C for the range of 1–30 dpa for most polycrystalline SiC materials, and (3) the effects on irradiation creep of initial microstructures—such as grain boundary, crystal orientation, and secondary phases—increase with increasing irradiation temperature.

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 °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 °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 °C, the ductility of K-doped W-3%Re after irradiation was significantly higher than that of Pure W. The irradiation at ~1000 °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 ~1000 °C, and ductility was exhibited without an increase in strength.

In this study, energetic helium (He) ion irradiation was performed to obtain bulk He distribution in tungsten (W) materials, concurrent with damage introduction at high temperature. Then, deuterium (D) implantation and thermal desorption spectrometry were performed to evaluate D retention. At the same time, the surface tritium (T) concentration and depth distribution were evaluated by imaging plate (IP) and β-ray induced X-ray spectroscopy (BIXS) measurements after mixed D-T gas exposure. Numerical simulations were applied to evaluate changes in binding energies, diffusion depths, and trapping sites under different irradiation conditions. The results showed that weak trapping sites with higher concentration, such as vacancies, were produced during only energetic He+ irradiation events, leading to enhancement of D retention. Fe3+-He+ simultaneous irradiation promoted the formation of HexVy complexes, which reduced the concentration of vacancy trapping sites and changed the stress field around defects, leading to the suppression of D trapping behavior. From the reduced effects of D retention caused by HexVy complexes at higher temperatures, the results suggested that defect recovery was the dominant mechanism. With increasing damage level at higher temperatures, more weak trapping sites, such as dislocations and vacancies sites, were produced, leading to a more dominant influence on D retention than HexVy complex effects. It was also found that HexVy complexes prevented D diffusion to the bulk and that simulation results showed that the damage level had little impact on D diffusion depth.

Neutron irradiation effects on the tensile properties of Pure W, K-doped W, W–3%Re, and K-doped W–3%Re were examined under the US-Japan collaboration project PHENIX. The fission neutron irradiation experiments were carried out up to 0.74 dpa at 600 °C and 800 °C. Pure W (SR) showed irradiation hardening and loss of ductility after irradiation at 600 °C and 800 °C. K-doped W–3%Re (SR) also exhibited irradiation hardening but was ductile after irradiation. Characteristic points of the K-doped W–3%Re (SR) are small grain size and layered structure. The stress-relief treatment and the layered structure may improve the ductility of powder-metallurgy W alloys after neutron irradiation, so a combination of K-doping and Re addition would be beneficial to improve irradiation resistance.

The hydrothermal corrosion behavior (320 °C, 20 MPa, 168 h) of high-purity chemical-vapor-deposited (CVD) SiC pre-irradiated with 5.1-MeV Si ions at 400 and 800 °C and 0.1–2.6 dpa was studied in order to clarify the effects of irradiation damage on SiC corrosion. Regardless of the pre-irradiation conditions, selective corrosion was observed at the grain boundaries and stacking faults even at the unirradiated regions. In contrast to the complete loss of the irradiated regions observed in the specimens irradiated at 400 °C during the autoclave test, a number of large grains survived in the case of the specimens irradiated at 800 °C. The corrosion rates at the irradiated regions increased with increasing irradiation fluence, with a significant dependence in the lower dpa regime similar to that observed in the point-defect swelling. SiO 2 formation was not detected in any case. Cross-sectional scanning transmission electron microscopy (TEM) and electron energy loss spectroscopy (EELS) analyses of the surfaces of the surviving grains revealed oxygen diffusion to a depth of 3.0 nm from the surface. A significant reduction of the oxygen diffusion barrier at the surface was implicated as one of the key mechanisms of the acceleration of the ion-irradiated SiC corrosion rates.

W samples were irradiated at 300 K with 6 MeV Fe ion with damage concentrations in the range from 0.0003 to 1.0 displacements per atom (dpa) and then implanted at 300 K with 500 eV D ions to a fluence of 5 × 10 21 D/m 2 . Deuterium retention in the damaged samples was examined in situ by thermal desorption spectrometry (TDS). Simulation of the TDS spectra was performed using the Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code to reveal the binding energies for deuterium captured by the ion-induced defects. It has been shown that the deuterium TDS spectra consist of two or three peaks (depending on the damage concentration) at about 400, 600 and 800 K, and can be simulated by the HIDT simulation code with the use of hydrogen-trap binding energies of 0.65, 1.25, and 1.55 eV.

Composites consisting of a SiC matrix prepared by chemical vapor infiltration that was reinforced by Tyranno™ SA 3rd (TSA3) or Hi-Nicalon™ Type S (HNLS) SiC fibers were irradiated with 5.1 MeV Si ions to 100 dpa at 300 °C. It was confirmed that the irradiated microstructure of TSA3 was stable up to 100 dpa. In contrast, the results of surface profilometry and cross-sectional transmission electron microscopy (TEM) showed that HNLS fibers underwent 0.8% shrinkage along the axis and 0.7% radial shrinkage. High-resolution TEM images of the highly damaged regions in the HNLS material revealed the complete loss of carbon ribbons initially distributed at the SiC grain boundaries. Ion-beam-induced diffusion of carbon into the SiC grains was indicated from the observations of the interdiffusion layer formed at the fiber/pyrocarbon interface. The π ∗ peak of the carbon K-edge spectrum was found in the electron-energy-loss spectrum of the HNLS SiC grains that were irradiated above 60 dpa; this peak was not observed in unirradiated SiC fibers, further demonstrating that carbon sp 2 bonding was only detected in HNLS SiC fibers. According to the changes in the bonding and the volume relaxation related to the carbon defects in SiC, an increase in the population of antisite C Si was likely the key underlying mechanism of the irradiation-induced shrinkage of the HNLS fibers.

The irradiation creep behavior of nano-powder sintered silicon carbide was investigated using the bend stress relaxation method under neutron irradiation up to 1.9 dpa. The creep deformation was observed at all temperatures ranging from 380 to 1180 °C mainly from the irradiation creep but with the increasing contributions from the thermal creep at higher temperatures. The apparent stress exponent of the irradiation creep slightly exceeded unity, and instantaneous creep coefficient at 380–790 °C was estimated to be ∼1 × 10 −5 [MPa −1 dpa −1 ] at ∼0.1 dpa and 1 × 10 −7 to 1 × 10 −6 [MPa −1 dpa −1 ] at ∼1 dpa. The irradiation creep strain appeared greater than that for the high purity SiC. Microstructural observation and data analysis indicated that the grain-boundary sliding associated with the secondary phases contributes to the irradiation creep at 380–790 °C to 0.01–0.11 dpa.

The silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite system for fusion applications has seen a continual evolution from development a fundamental understanding of the material system and its behavior in a hostile irradiation environment to the current effort which is directed at a broad-based program of technology maturation program. In essence, over the past few decades this material system has steadily moved from a laboratory curiosity to an engineering material, both for fusion structural applications and other high performance application such as aerospace. This paper outlines the recent international scientific and technological achievements towards the development of SiC/SiC composite material technologies for fusion application and discusses future research directions. It also reviews the materials system in the larger context of progress to maturity as an engineering material for both the larger nuclear community and broader engineering applications.

Silicon carbide (SiC) for nuclear structural applications, whether in the monolithic ceramic or composite form, will require a robust joining technology capable of withstanding the harsh nuclear environment. This paper presents significant progress made towards identifying and processing irradiation-tolerant joining methods for nuclear-grade SiC. In doing so, a standardized methodology for carrying out joint testing has been established consistent with the small volume samples mandated by neutron irradiation testing. Candidate joining technologies were limited to those that provide low induced radioactivity and included titanium diffusion bonding, Ti–Si–C MAX-phase joining, calcia–alumina glass–ceramic joining, and transient eutectic-phase SiC joining. Samples of these joints were irradiated in the Oak Ridge National Laboratory High Flux Isotope Reactor at 500 or 800 °C, and their microstructure and mechanical properties were compared to pre-irradiation conditions. Within the limitations of statistics, all joining methodologies presented retained their joint mechanical strength to ∼3 dpa at 500 °C, thus indicating the first results obtained on irradiation-stable SiC joints. Under the more aggressive irradiation conditions (800 °C, ∼5 dpa), some joint materials exhibited significant irradiation-induced microstructural evolution; however, the effect of irradiation on joint strength appeared rather limited.

Irradiation-induced creep in high-purity silicon carbide was studied by an ion-irradiation method under various irradiation conditions. The tensioned surfaces of bent thin specimens were irradiated with 5.1 MeV Si 2+ ions up to 3 dpa at 280–1200 °C, which is referred to as a single-ion experiment. Additional He + ions were irradiated simultaneously in the dual-ion experiment to study the effects of transmuted helium on irradiation creep. Irradiation creep was observed above 400 °C in the single-ion case, where a linear relationship between irradiation creep and swelling (C/S) was observed at 400–800 °C for all stress levels (150, 225, and 300 MPa). The proportional constant of the C/S relationship was strongly dependent on temperature and stress. A rapid reduction in creep strain was observed above 1000 °C. On the basis of the microstructural analysis, anisotropic distribution of self-interstitial atom (SIA) clusters was suspected to be the primary creep mechanism. Some interesting results were obtained from re-irradiation under stress after the irradiation without stress. The creep strain was significantly retarded by pre-irradiation to even 0.01 dpa at 400 and 600 °C. This implies that the loop orientation was determined very early in the irradiation regime. For the dual-ion cases, irradiation creep was absent or very limited at all irradiation temperatures studied (400–800 °C). Microstructural analysis indicated that helium inhibited the stable growth of SIA clusters and prevented them from exhibiting anisotropic distribution.

Unidirectional silicon carbide (SiC)-fiber-reinforced SiC matrix (SiC/SiC) composites fabricated by a nano-infiltration and transient eutectic-phase (NITE) process were irradiated with neutrons at 600 °C to 0.52 dpa, at 830 °C to 5.9 dpa, and at 1270 °C to 5.8 dpa. The in-plane and trans-thickness tensile and the inter-laminar shear properties were evaluated at ambient temperature. The mechanical characteristics, including the quasi-ductile behavior, the proportional limit stress, and the ultimate tensile strength, were retained subsequent to irradiation. Analysis of the stress–strain hysteresis loop indicated the increased fiber/matrix interface friction and the decreased residual stresses. The inter-laminar shear strength exhibited a significant decrease following irradiation.

The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C β-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ∼3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation.

Mechanical properties of silicon carbide (SiC)-fiber-reinforced SiC matrix (SiC/SiC) composites are sensitive to residual stresses induced during the fabrication process. Differential swelling between individual SiC fibers and the SiC matrix due to irradiation modifies the residual stresses, thereby affecting the strength of the irradiated composites. To understand the effect of irradiation on the strength of SiC/SiC composites, the swelling of the fibers and matrix was evaluated following 5.1 MeV Si2+ ion irradiation corresponding to a dose of ∼3 dpa at 873–1273 K. The magnitude of swelling in Tyranno™-SA3 SiC fiber and a chemical-vapor-infiltrated SiC matrix was similar. The magnitude of swelling of the matrix was larger than that of the SA3 fibers by 0.2–0.5% in nano-infiltrated and transient eutectic phase (NITE) SiC/SiC composites at 873–1273 K to 3 dpa. This swelling mismatch can reduce the residual tensile stress in the NITE-SiC matrix and in turn improve the proportional limit stress of NITE SiC/SiC composites after neutron irradiation.

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.

This paper reviews recent achievements as to "nuclear-grade" SiC composites in particular for materials-system integration. SiC composite component development are reviewed including VHTR control rod scale model and compact intermediate heat exchanger scale mode by current joining and assembly techniques. Joining methods for SiC to metal and results of characterization of joint shear strength by the torsion tests using small specimens were also reviewed. The recent results of neutron irradiation experiments were also reviewed including detailed analysis of mechanical properties, irradiation creep and preliminary results on tritium behavior in SiC.

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.

Silicon carbide (SiC) samples have been joined by a pressureless slurry based method. The CaO–Al2O3 (CA) glass-ceramic joining material has been characterized in term of crystalline phases and thermal and mechanical properties. A torsion test based on miniaturized hourglass shaped specimens has been used as pure shear strength test method for joined ceramic samples. Torsion results are compared to those obtained by a single lap offset (SLO) test in compression on the same joined materials. Pure shear strength of 104 ± 25 MPa has been measured by torsion test, whereas single lap offset gave 36 ± 8 MPa.