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High-melting-point borides are attractive for structural and functional materials used in extreme environments. However, their poor sinterability due to their high melting point and strong covalent bonding have prevented their further applications. In this study, a new sintering process, called transient liquid-phase diffusion sintering (TLPDS), has been developed for high-melting-point borides. TiB2 compacts were fabricated by spark plasma sintering at 50 MPa for 15 min at temperatures of 1300 and 1600 °C via TLPDS of TiB2 powder with a eutectic TiB–Ti powder as the sintering aid, prepared via mechanical alloying (MA). Differential scanning calorimetry results indicated that the melting temperature of the obtained sintering aid was lower than its equilibrium melting point. MA of the sintering aid suppressed the open pore fraction to one-third of that in the compact sintered with the aid prepared without MA when sintering at 1300 °C. We also propose a possible mechanism for TLPDS of TiB2.

The relationship between lattice defects and the corrosion of SiC irradiated with heavy ions at 400 °C and 800 °C was investigated using electrochemistry and electron spin resonance spectroscopy. An immersion test in high-pressure hot water was also used to evaluate surface recession. The corrosion current increased monotonically with dangling-bond (DB) density, indicating that DBs were the primary reaction sites. Further, the surface recession rate in hot water increased exponentially with increasing DB density. The dense DBs were found to be surrounded by distorted lattices. These findings indicated that SiC dissolution was further accelerated by the accumulated lattice distortion.

Lithium metal and lithium oxides are components of lithium–oxygen (Li–O2) batteries. To accurately identify Li compounds and understand the degradation mechanism, fundamental knowledge of the elect...

The use of an electronic descriptor is a high-throughput approach for predicting the chemical reactivities of various materials that has the potential to significantly facilitate the material design of Be intermetallic neutron multipliers for nuclear fusion applications. However, electronic descriptors for Be intermetallics with the desired properties, including radiation tolerance and reduced hydrogen retention, have been poorly understood. Herein, we perform first-principles calculations on 42 existing binary Be intermetallics to find an effective electronic descriptor. We demonstrate that the occupied Be p band center relative to the Fermi level is a bulk descriptor, correlating with the Be vacancy formation energy; a positive and linear correlation with a coefficient of determination R² = 0.85 was observed for Be₁₂X (X: transition metal). The upward shift in energy of the occupied Be p states with early or middle transition metals can reduce hydrogen solution energy, which could be attributed to the less filled anti-bonding state of interstitial hydrogen atom. It is confirmed that the bulk descriptor is an experimentally measurable scale, having the strong linearity (R² = 0.97) with the calculated; therefore, it can accelerate the material development of beryllium intermetallic neutron multipliers., 軽元素材料の特性を予測する電子的記述子を発見 --情報駆動型の材料開発に道筋--. 京都大学プレスリリース. 2022-10-21.

The phase transition of Fe1-xCrx alloys after thermal aging at 748 K has been investigated by means of positron annihilation spectroscopy. It was observed that the implanted positrons preferentiall...

Static discontinuous recrystallization was studied during room temperature annealing of a newly designed 80% room temperature rolled oxide dispersion strengthened copper. At early stages of annealing, fine new oriented nanosized/submicron grains were recrystallized in the unique matrix of single brass-oriented deformed grain. Upon longer annealing time up to 14 months, the size, area fraction and number density of the recrystallized grains increased significantly along with changing the crystallographic textures. The analysis of misorientation angle distribution of boundaries indicated transformation of low angle boundaries to high angle boundaries results in nucleation of recrystallized grains by significant contribution of static recovery. Furthermore, the constant level of mechanical hardness after recrystallization was interpreted by the balance between grain size hardening, oxide particle hardening and strain hardening. (C) 2019 Elsevier B.V. All rights reserved.

Currently, advanced ODS copper alloy is under study as a potential fusion material providing good mechanical properties. In this work, in order to develop a high performance ODS copper containing 0.5 wt% Y2O3 oxide particles, the effect of room temperature rolling and subsequent annealing on the grain structure evolution, texture development and tensile properties are studied using EBSD, TEM and tensile tests. Microstructure evolution studies show the grain structure coarsens by enhancing the Brass texture during increase of rolling reduction and a unique single crystal-like brass-texture deformed structure is achieved after 80% rolling reduction. We found the deformation mechanism of partial slip by a 6 〈 211 〉 dislocations facilitated by the pinning of a 2 〈 101 〉 perfect dislocations through fine oxide particles is responsible for formation of Brass texture during room temperature rolling. Furthermore, the recrystallization of ODS copper retards to high temperature of ~700 °C and shows a fine-grained microstructure with different orientations of Goss, Brass, S and Copper. Evaluation of microstructure-mechanical properties of the recrystallized samples expresses that the bimodal grain size distribution at 800 °C for 30 min offers a good tensile strength-ductility (UTS: 491 MPa, elt: 19%) at ambient temperature.

Be-rich intermetallics (beryllides) have gathered wide attentions to be adopted for high temperature environments, such as an advanced neutron multiplier in fusion reactors. This study reveals the ...

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.

The effect of hydrogen on the hardness and tensile properties of pure tungsten was examined using Vickers hardness and tensile tests. Samples were exposed to high-pressure hydrogen gas (5.8 MPa). The tensile behavior, tensile fracture surface, and hardness of as-received and stress-relieved tungsten did not change after hydrogen charging, owing to the low solubility of hydrogen. Therefore, to understand the effect of hydrogen on these materials, experiments must be performed to trap more hydrogen atoms at dislocations. In contrast, the hardness of electron-irradiated tungsten increased after hydrogen charging. Additionally, after a heat treatment at 473 K, hydrogen atoms dissociated from single vacancies, and the hardness decreased to the pre-charged value. Thus, single vacancies decorated with hydrogen atoms are expected to obstruct dislocation motion. Keywords: Tungsten, Hydrogen, Hardness, Tensile test, Defects, Electron irradiation

The lead-lithium (Pb-Li) alloy is considered as a coolant and a tritium breeder for fusion reactor blanket systems. One of the critical requirements in realization of the systems is the compatibility of silicon carbide (SiC) and its composites, as structural and/or functional materials of the blanket, with liquid Pb-Li flow at high temperatures. This study aims to clarify the corrosion process of high-purity SiC materials under flow condition of liquid Pb-Li for 3000 h at 700 and 900 °C, respectively, by using the rotating disc corrosion equipment. Furthermore, the effect of wall flow velocity on the compatibility was specifically evaluated. For that purpose the decrease rate (about 32–37%) of the wall flow velocity on disc type’s sample by the dragging effect by fluidic viscosity was first calculated by numerical analysis of the rotating flow condition. The wall flow velocity dependence on the reaction layer’s thickness was then evaluated for both high purity SiC and SiC/SiC composites. The test results show that the formation process of the surface reaction layer mainly depends on both the time required for the chemical reaction and the wall flow velocity in the initial stage but the reaction layer thickness tended to approach to the constant regardless of the wall flow velocity with increasing exposed time. In short, no accelerated corrosion was found for about 3000 h. It was also found that there was no significant difference of the temperatures between 700 and 900 °C. Finally the key corrosion mechanism, i.e., the reaction of oxide impurities, e.g., Li2O, in liquid Pb-Li with the Si-oxide layer over SiC, was identified.

The feasibility of liquid metal atomization in vacuum is studied by analysis and verified by experiment. Simultaneous tritium and heat recovery in vacuum from liquid lithium–lead (PbLi) has been previously proposed, where hot and tritium–rich PbLi is transformed into small droplets in vacuum, with the tritium then released by advection and the heat radiated away. Both mechanisms aimed to reduce the tritium permeation and membrane corrosion. The key hurdle is the atomization of liquid metal in vacuum by a large diameter nozzle. Conventional sprays function in air or other gases but not in vacuum. Therefore, an instability is applied that occurs at the velocity inflection in a parallel flow. The velocity inflection is generated by superimposing rotation on a liquid jet with a dedicated swirl nozzle. This can function in vacuum. An experiment using a low–temperature melting alloy (GaInSn) is performed in vacuum, using nozzle diameters of 0.7 mm, swirl angles 67°, and flow velocities of 3 and 5 ms−1. The instability that arises when using GaInSn at room temperature resembles that of PbLi at 400–500 °C, can be used as a proxy. The observed mean droplet Feret diameter was 0.31 mm at 3 m/s jet velocity and 0.23 mm at 5 m/s. These results support the feasibility of small droplet formation from a large diameter nozzle in vacuum. Further experiment must be performed to fully verify the droplet formation and proposed tritium and heat extraction method.

Radioactivity levels for conventional thermal diffusivity test specimens are often prohibitively high for examination with candidate metallic materials, such as tungsten and reduced activation steels for fusion energy. Therefore, in the ongoing Japan–US PHENIX project, irradiation in the high flux isotope reactor (HFIR) was performed with the miniature specimen form of diameter (D) 3 mm × thickness (T) 1/2 mm small disk for thermal diffusivity measurement. In this study, the thermal diffusivity measurement of this miniature specimen is validated using a Netzsch LFA-467 ‘Hyper Flash’ thermal analyzer and ‘graphene nanoplatelets containing agent’. With careful surface treatment and data analysis, the miniature specimens showed almost the same value as the diameter 10 mm standard specimens.

An action plan presented here is the plan toward construction of a Demo reactor in Japan. A Task-force for development strategy of Tokamak Demo Reactors was established and has considered an action plan to construct the Demo in the 2040s. The action plan will lead works in a Joint Special Design Team for Fusion Demo established for design and R&D of the Demo reactor in Japan. Although the first version of action plan was reported in March 2016, a revised version, presented here, is being drafted in order to correspond to the new schedule of the ITER project agreed in 2016, because the ITER progress is one of critical issues in the action plan.

The fusion-biomass hybrid model system, which takes waste biomass from municipal and agricultural areas as well as forests as feedstock, produces either diesel through the Fischer–Tropsch (FT) reaction or electricity by the solid oxide fuel cell (SOFC). This system produces synthesis gas by endothermic pyrolytic gasification using high temperature fusion heat. A temperature of over 700 °C of exterior thermal heat from the fusion reactors with a duel cooled lithium lead blanket and its technical extension bring about biomass gasification to produce maximum amounts of chemical energy and synthetic gas, from feedstock. First, synthetic gas that contains hydrogen (H2) and carbon monoxide (CO) can be converted into artificial diesel (–CH2–), which is regarded as “carbon–neutral”. The other is to generate electricity by putting synthetic gas into SOFC at various scales, not only at the plant scale but also at the residential scale. This paper aims to conduct an economic analysis of the fusion-biomass hybrid model by comparing diesel and SOFC electricity under the assumption of the investment of a biomass plant with an FT reaction facility and one with SOFC. A sensitivity analysis is performed applying diesel price, electricity price, SOFC efficiency, diesel subsidy, and fusion heat cost. These results can help in targeting which products are economically justified in the circumstances of variable environmental policies under different policies and economic situations, which would have a significant impact on commercial fusion designing.

Yttria (Y2O3) dispersed copper alloys were fabricated by an in-situ process combining Mechanical Alloying (MA) and Hot Isostatic Pressing (HIP). In this process, CuO powder was added into Cu(Y) alloy powders halfway during mechanical alloying (MA). To investigate the effects of Y + CuO amount on the MA process and HIP process respectively, four kinds of samples (Cu-0/1/3/5wt%Y2O3) were studied in this research. For powders after MA, it was confirmed by an X-ray diffraction system (XRD) that more Y + CuO addition can refine the grain size. By calculating the lattice parameters from the XRD data, and analyzing the thermal dynamics data that were obtained from the TG-DTA tests, it can be concluded that the Cu, Y were alloyed successfully at first, and then after the CuO addition, reaction between Cu(Y) and CuO occurred, extracting part of Y solute from the Cu(Y) solution. Consequently, the alloying and grain refinement effect together enhance the microhardness of the powders. After consolidation by HIP, the formation of Y2O3 particles were confirmed by XRD. The relative density and the Vickers hardness of the HIP samples increased with the increasing Y2O3 amounts, whereas the electrical conductivity showed an opposite change tendency. Keywords: Copper alloy, Dispersion strengthening, Copper oxide, Mechanical alloying, Hot isostatic pressing

The authors constructed a dynamic simulation model of a nuclear fusion power plant on Modelica language to obtain fundamental knowledge on the plasma control requirements for the future commercial fusion power plants. The fusion power plant model was designed with a 1500-MW thermal output tokamak reactor with He-cooled Li2TiO3 solid breeder blanket (coolant outlet conditions: 8 MPa and 515.8 °C). A superheated Rankine cycle was designed to achieve the electrical output of 485.38 MW with the operating pressure of 20.5 MPa. Two plasma output patterns, a step decrease of power and a single pulse decrease of power, were simulated to assess the response of the power plant. A sudden step decrease in fusion neutron led to an immediate decrease in the blanket temperature and the first coolant temperature. In order to avoid the sharp temperature drop, a need for a turbine bypass mechanism or a He coolant boiler bypass mechanism was indicated. On the other hand, because of the delay in the plant responses, the deviation of the electrical output from steady state could be minimized by recovering the plasma output in few tens of seconds. Based on the findings, a new diagram was presented that illustrates an important plasma control requirements for future commercial fusion power plants.

Phase separation in Fe–Cr binary alloys irradiated with neutrons at 473 K and 573 K was investigated using positron annihilation spectroscopy. Using positron annihilation coincidence Doppler broadening (CDB) measurements, the phase separation progress was observed in neutron-irradiated samples at 473 K and 573 K. Vacancy clusters were detected in Fe–xCr (x = 0, 9, 15, 30, 45, 50, and 100) during 473 K irradiation using positron annihilation lifetime measurements, but were not detected in Fe–xCr (x = 70, 85, and 91) irradiated at 473 K or in any samples irradiated at 573 K. Additionally, in Fe–xCr (x = 70, 85, and 91) irradiated at 473 K, all positrons were annihilated with core Fe electrons as determined from CDB ratio curves. Thus, vacancy clusters were not detected in the Fe-rich phase. There was a possibility that vacancy clusters are formed in the Cr-rich phase, but they were not detected by the PAS. Therefore, another method is necessary to investigate this further. Vickers hardness tests indicated that neutron-irradiated samples were harder than unirradiated samples. The contribution of phase separation and neutron-irradiation defects to increased hardness was dependent on the irradiation conditions including temperature and dose. Keywords: Fe–Cr binary alloys, Neutron irradiation, Positron annihilation, Hardness, Phase separation

Advanced oxide dispersion strengthened copper alloys are promising structural materials for application in divertor system of future fusion reactors due to high irradiation resistance, high thermal conductivity, and good mechanical properties. In this study, a new ODS copper including 0.42wt%Y2O3 nanosized oxide particles was developed successfully by mechanical alloying method using addition of 1 wt% Stearic acid in Ar atmosphere. Mechanical alloying resulted in decrease of crystallite size to 28 nm in concurrent with increment of dislocation density and hardness to the saturated level of 1.7×1015m−2 and 226HV0.1 after 48 h milling, respectively. Consolidated ODS copper by SPS and then hot roll-annealing at 900 °C/60 min showed an average grain size of 1.1 µm with a near random texture. Furthermore, TEM observations demonstrated fine semicoherent Y2O3 oxide particles distributed with a misfit parameter (δ) of 0.17 in copper matrix with an average size of 10.8 nm and interparticle spacing of 152 nm. Finally, tensile test evaluation determined comparable mechanical properties of the annealed ODS copper (Cu-0.42wt%Y2O3) with Glidcop-Al25 including a yield strength of 272 MPa and total elongation of 12%, by two mechanisms of grain boundary strengthening and oxide particle strengthening. Keywords: ODS copper, Mechanical alloying, Microstructure, Mechanical properties

ODS ferritic steels were produced by mechanical alloying (MA) in either attritor or planetary ball mills under argon atmosphere and hot extrusion. Milling in argon with controlling excess oxygen (Ex.O) and nitrogen was effective to reduce the particle size of MA powders. Coarse Al2O3 particles and fine AlN precipitates were found to be formed in an Al added steel when the Ex.O and nitrogen concentrations are increased. Lowering the Ex.O and nitrogen concentrations in ODS ferritic steels acted as a key parameter for enhancing their impact properties.

The interface bonding strength of thinly coated materials is difficult to be measured accurately owing to the size limitations of the coating layer. In this study, the interfaces bonding strength of an underwater explosive welded tungsten foil on ferritic steel was investigated by deep-notch-type micro-tensile tests and microscopic double-notch shear (DNS) compression tests. The micro-tensile tests were carried out using a focused ion beam (FIB) tensile system. The results showed that the tensile strength of the interface region near the ferritic steel was larger than that of the ferritic steel but weaker than that of the tungsten foil. The micro-DNS compression tests were conducted in both the tungsten and interface regions. The obtained shear strength of tungsten exhibited a strong size dependence as follows: the smaller the size, the higher the strength. Although interface specimens with a large gauge section area showed a shear strength similar to that of tungsten DNS specimens, an interface region specimen with an area of the gauge section of approximately 1 μm ² showed larger shear strengths (∼4.8 GPa) than the tungsten region (3.2 GPa). Observations of the bonding microstructure and fracture surface indicated that the obtained shear strength with 1 μm ² specimens corresponded to the interface bonding shear strength between tungsten and ferritic steel obtained by underwater explosive welding.

The small specimen test technique and the less-specimen-consumed test method for evaluating the fracture toughness are highly required in the nuclear field. In the present study, the fracture toughness properties of the F82H steels with the size of 1/4 CT and 1/6 CT were evaluated by the master-curve approach. Comparing the 1/4 CT and 1/6 CT data with that obtained by the 1 CT specimen, it shows no substantial effect of the specimen size on the evaluation of the fracture toughness by the master-curve approach.

Any fusion power reactors such as “DEMO” that achieves tritium self-sufficiency with breeding blankets can produce tritium by DD reaction followed by exponential breeding in the blanket within reasonable total operation period. The present study further suggests that realistic Power Ascension Tests (PAT) of DEMO can produce its tritium to be needed in the series of tests by its own program until reaching steady state full power operation, with no external supply or additional operation costs. Closed tritium fuel plant was described by a system dynamics model, and analyzed considering realistic PATs of DEMO, that will be mainly pulsed DD and low concentration DT. Typical PATs require years of operation from zero power criticality to full power, with pulsed power output and long dwell time between them. Output power is gradually increased in PATs to check the functions of reactor systems and components. In the case of fusion DEMO, zero power criticality corresponds to DD operation. While plasma may be fired in pulses, tritium plant is continuously operated to recover all the tritium produced by the DD and low DT burn. Depending on the different time constant of tritium retention in components, tritium is transferred by deuterium purge, and high concentration tritium is finally collected in the storage, to be available for the next tests at virtually no additional cost. Realistic “initial loading tritium” will be 100 g for commissioning of tritium plant far prior to the plasma operation.