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The self-guided molecular dynamics (SGMD) method, which can enhance the conformational sampling efficiency in MD simulations, was applied in investigating the phase transformation of Cu precipitate in α-iron that took place during thermal aging. It was shown that the SGMD method can accelerate calculating the bcc to 9R structure transformation of a small precipitate (even 4.0 nm in size), enabling the transformation without introducing any excess vacancies. The size dependence of the transformation also agreed with that seen in previous experimental studies. [ABSTRACT FROM AUTHOR]

A combination of simulated annealing with Langevin molecular dynamics and the basin-hopping with occasional jumping (BHOJ) technique was used to systematically determine the most stable configurations of self-interstitial atom (SIA) clusters In (n = 1-38) in α-iron. In addition to the original BHOJ technique, we introduced an additional long jumping process in which a randomly selected less-bounded atom is moved to a neighbouring site of another SIA in the cluster to enhance the probability of locating the global minimum structure. With the obtained putative lowest energy structures, the binding energies as a function of cluster size were estimated. We also determined the sizes of particular stable clusters based on their geometrical symmetry. Furthermore, the values were extrapolated based on accurately determined formation energies, and are available for immediate use in kinetic Monte Carlo or rate theory models. [ABSTRACT FROM AUTHOR]

Abstract: A reduced activation ferritic steel and an austenitic stainless steel were irradiated at temperatures of 230°C and 350°C in an irradiation capsule with temperature control capability independent from reactor power to accumulated damage levels of about 1.5 displacement per atom (dpa). For some of the specimens temperature was changed during irradiation. The temperature change reduced the irradiation hardening of the austenitic steel. Conversely, it slightly increased the hardening of the reduced activation ferritic steel. The mechanism of the observed temperature change effect and the impact of the additional hardening on the residual ductility is discussed. [Copyright &y& Elsevier]

Abstract: The dependence of helium production on radiation hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel (8Cr–2W–0.2V–0.04Ta–0.1C) irradiated at 250°C to 2.3dpa. In this study, 10B and 11B-doped specimens were irradiated to minimize the errors from the effect of B on mechanical properties by comparing the results. The specimens used were 10B-doped, 10B+ 11B-doped and 11B-doped F82H steels. The total amounts of doping boron were about 60 mass ppm. The range of helium concentration produced in the specimens was from about 5 to about 330appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50MeV-He2+ irradiation was also performed to implant about 85appm He atoms at 120°C by AVF cyclotron to 0.03dpa, and small punch testing was performed to obtain ductile-to-brittle transition temperatures (DBTT). Radiation hardening of the neutron-irradiated specimens increased slightly with increasing helium production. The 100MPam1/2 DBTT for the F82H+ 11B, F82H+ 10B+ 11B, and F82H+ 10B specimens were 40, 110, and 155°C, respectively. The shifts of DBTT due to helium production were evaluated as about 70°C by 190appm He and 115°C by 330appm He. In cyclotron experiment using standard F82H, a similar DBTT shift due to He was measured. These results suggest that helium production can increase the DBTT. [Copyright &y& Elsevier]

Abstract: F82H (Fe–8Cr–2W) and its variant doped with 2%Ni were irradiated up to 20dpa at 300°C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load–displacement curves and the strain distributions obtains from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300°C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400°C. [Copyright &y& Elsevier]

Significant progress has been achieved in the international research effort on reduced-activation steels. Extensive tensile, fracture toughness, fatigue, and creep properties in unirradiated and irradiated conditions have been performed and evaluated. Since it is not possible to include all work in this limited review, selected areas will be presented to indicate the scope and progress of recent international efforts. These include (1) results from mechanical properties studies that have been combined in databases to determine materials design limits for the preliminary design of an ITER blanket module. (2) Results indicate that the effect of transmutation-produced helium on fracture toughness is smaller than indicated previously. (3) Further efforts to reduce irradiation-induced degradation of fracture toughness. (4) The introduction of a post-irradiation constitutive equation for plastic deformation. (5) The production of ODS steels that have been used to improve high-temperature strength. (6) The method developed to improve fracture toughness of ODS steels. [Copyright &y& Elsevier]

Abstract: Breeding blankets are the most important components in DEMO. The DEMO blanket has to withstand high neutron flux typically 15–30 dpa/year under continuous operation. Therefore integrated and effective development of blanket structural materials and breeding/multiplying materials is essential in the blanket development for DEMO. In parallel to the ITER program, broader approach (BA) activities are initiated by EU and Japan. Based on the common interest of each party towards DEMO, R&D on reduced activation ferritic martensitic (RAFM) steels as a DEMO blanket structural material, SiCf/SiC composites which have potential for use in DEMO blankets, advanced tritium breeders and neutron multiplier for DEMO blankets, and tritium technologies including tritium behavior studies in advanced materials for DEMO blanket applications will be carried out as a part of the BA activities. [Copyright &y& Elsevier]

Abstract: Recent research results obtained in Europe, Japan, China and the USA on reduced activation ferritic/martensitic (RAFM) steels are reviewed. The present status of different RAFM steel products (plate, powder HIPped steel, many types of fusion and diffusion welds, unirradiated and irradiated states) is sufficient to present a strong case for the use of the steels in ITER test blanket modules. For application in DEMO, more research is needed, including the use of the International Fusion Materials Irradiation Facility (IFMIF) in order to quantify the effects of large amounts of transmutation products, such as helium and hydrogen. [Copyright &y& Elsevier]

Abstract: The effect of implanted helium (He) on the thermal diffusivities of SiC/SiC composites was investigated. In the result, thermal diffusivities of SiC/SiC composites decreased after He implantation. The thermal diffusivities of implanted specimens were partly recovered by annealing. From the obtained results in this study, the defect concentration induced by He implantation in the specimens was estimated. The defect concentration rapidly decreased around 500°C. The reason is that He release from SiC starts at 500°C. The defect concentration induced by He implantation increased with increasing the amount of implanted He in the He implantation range less than 30appm He. [Copyright &y& Elsevier]

A Japanese ITER test blanket module (TBM) is planed to use reduced-activation martensitic steel F82H. Feasibility of F82H for ITER test blanket module is discussed in this paper. Several kinds of property data, including physical properties, magnetic properties, mechanical properties and neutron-irradiation data on F82H have been obtained, and these data are complied into a database to be used for the designing of the ITER TBM. Currently obtained data suggests F82H will not have serious problems for ITER TBM. Optimization of F82H improves the induced activity, toughness and HIP resistance. Furthermore, modified F82H is resistant to temperature instability during material production. [Copyright &y& Elsevier]

Abstract: As a part of the Broader Approach activities, R&D on blanket related materials, reduced-activation ferritic martensitic (RAFM) steels as a structural material, SiCf/SiC composites for flow channel insert in the liquid blanket and/or use as advanced structural material, advanced tritium breeders and neutron multiplier, has been initiated directed at DEMO. As part of the RAFM steel mass production development, a 5ton heat of RAFM steel (F82H) was procured by Electro Slag Re-melting as the secondary melting method, which was effective in controlling unwanted impurities. An 11ton heat of EUROFER was also produced. For the SiCf/SiC composite development, NITE- and CVI-SiCf/SiC composites were prepared as reference materials and preliminary mechanical and physical properties were measured. Also compatibility tests between SiC and Pb–17Li have been prepared, related to the He-cooled Li–Pb blanket concept. For the beryllide neutron multiplayer Be–Ti alloy development, large size rods of about 30mm diameter were fabricated successfully in EU. [Copyright &y& Elsevier]

Abstract: Embrittlement is known to be caused by P segregation at grain boundaries in Fe alloys. Effects of P substitutions on binding energies and electronic structures of octahedral Fe cluster are investigated using density functional calculations in order to understand the nature of bonding between P and Fe atoms at grain boundaries. The binding energies increase in Fe3P3 and Fe-rich clusters while they decrease in P-rich clusters. The changes in binding energies are closely connected to the charge transfer from Fe to P atoms. The charge transfer leads to both stronger and weaker bonds in mixed Fe–P clusters. The weaker bonds due to less charge cause embrittlement. The calculations indicate that the binding energies and chemical bonding are affected by atomic configurations of P atoms in Fe–P clusters. [Copyright &y& Elsevier]

Abstract: The radiation-hardening of oxide dispersion strengthened (ODS) alloys was examined using ion irradiation and nano-indentation. In this work, three ODS steels were irradiated in the TIARA facility at JAEA with 10.5MeV Fe3+ ions up to a dose of 20dpa at 250 and 380°C. Micro-hardness measurements were carried out on the ion-irradiated specimens with ultra-low load indention. Micro-structures were investigated by transmission electron microscopy (TEM) to examine the contribution of various types of defects to the radiation-hardening. All three steels showed increases in the hardness after the ion irradiation, and F82H-ODS showed the lowest radiation-hardening, which suggests that F82H-ODS has the better radiation resistance. Small amounts of particle dissolution was also confirmed in all of the steels after the irradiation. [Copyright &y& Elsevier]

Abstract: The status and key issues of reduced activation ferritic/martensitic (RAFM) steels R&D are reviewed as the primary candidate structural material for fusion energy demonstration reactor blankets. This includes manufacturing technology, the as-fabricated and irradiates material database and joining technologies. The review indicated that the manufacturing technology, joining technology and database accumulation including irradiation data are ready for initial design activity, and also identifies various issues that remain to be solved for engineering design activity and qualification of the material for international fusion material irradiation facility (IFMIF) irradiation experiments that will validate the data base. [Copyright &y& Elsevier]

Abstract: This paper summarizes the proposals and findings of the IFMIF Specification Working Group established to update the users requirements and top level specifications for the facility. Special attention is given to the different roadmaps of fusion pathway towards power plants, of materials R&D and of facilities and their interactions. The materials development and validation activities on structural materials, blanket functional materials and non-metallic materials are analyzed and specific objectives and requirements to be implemented in IFMIF are proposed. Emphasis is made in additional potential validation activities that can be developed in IFMIF for ITER TBM qualification as well as for DEMO-oriented mock-up testing. [Copyright &y& Elsevier]

Abstract: Significant progress has been achieved in the international research effort on reduced activation ferritic/martensitic steels for fusion structural applications. Because this class of steels is the leading structural material for test blankets in ITER and future fusion power systems, the range of ongoing research activities is extremely broad. Since, it is not possible to discuss all relevant work in this brief review, the objective of this paper is to highlight significant issues that have received recent attention. These include: (1) efforts to measure and understand radiation-induced hardening and embrittlement at temperatures ⩽400°C, (2) experiments and modeling to characterize the effects of He on microstructural evolution and mechanical properties, (3) exploration of approaches for increasing the high-temperature (>550°C) creep resistance by introduction of a high-density of nanometer scale dispersoids or precipitates in the microstructure, (4) progress toward structural design criteria to account for loading conditions involving both creep and fatigue, and (5) development of nondestructive examination methods for flaw detection and evaluation. [Copyright &y& Elsevier]

Abstract: The effects of He on the fracture behavior of reduced-activation ferritic/martensitic steels, including oxide dispersion-strengthened (ODS) steels and F82H, was determined by characterizing the microstructural evolution in and fracture behavior of these steels after He implantation up to 1000appm at around 550°C. He implantation was carried out by a cyclotron with a beam of 50MeV α-particles. In the case of F82H, the ductile-to-brittle transition temperature (DBTT) increase induced by He implantation was about 70°C and the grain boundary fracture surface was only observed in the He-implanted area of all the ruptured specimens in brittle manner. By contrast, no DBTT shift or fracture mode change was observed in He-implanted 9Cr-ODS and 14Cr-ODS steels. Microstructural characterization suggested that the difference in the bubble formation behavior of F82H and ODS steels might be attributed to the grain boundary rupture of He-implanted F82H. [Copyright &y& Elsevier]

Abstract: Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R&D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed. [Copyright &y& Elsevier]

Abstract: This paper describes the latest design of liquid lithium (Li) target system in International Fusion Materials Irradiation Facility (IFMIF). IFMIF is an accelerator-driven intense neutron source for fusion reactor materials testing. Two 40MeV deuteron beams with a total current of 250mA strikes a liquid Li target circulating in a Li loop. The Li target is to provide a stable Li jet at a speed of 10–20m/s to handle an averaged heat flux of 1GW/m2. A concaved Li flow is applied to avoid Li boiling. A cold trap and two kinds of hot trap are applied to control impurities (T, 7Be, C, O, N) below permissible levels. A back wall made of RAFM is located in the most severe region of neutron irradiation (50dpa/year). System availability requires more than 95% during plant lifetime. EVEDA tasks including Li loop are now being performed under the Broader Approach. [Copyright &y& Elsevier]

Abstract: The possibility of reducing the concentration of 14C produced in irradiation of reduced activation ferritic steels (RAFS) intended as structural materials in first wall and blanket of fusion power plants was investigated. During RAFS irradiation 14C is mainly produced from the more abundant of the two isotopes of nitrogen present in the steel, namely 14N. The method proposed consists in increasing the enrichment of the other isotope (15N). One-dimension transport calculations show that for a typical DEMO blanket configuration an 15N enrichment from the natural value (0.37%) to 95% was sufficient to keep the end of life 14C concentration in the RAFS below the limit (3.7×107 Bq/kg) fixed by the Japanese Nuclear Safety Commission for qualifying it as a low level material (LLM) which can be disposed by shallow land burial. [Copyright &y& Elsevier]

Abstract: Li2TiO3 ceramics have been irradiated with xenon (Xe) ions in the energy range of 18–160MeV at ambient temperature. The effect of ion irradiation on the order in the atomic arrangement of Li2TiO3 crystal was examined by X-ray diffraction (XRD) and Raman spectroscopy. The results show that the atomic arrangement in Li2TiO3 is strongly disordered by the high energy Xe ion irradiation. The destruction of long-range order due to the irradiation was confirmed from the reductions in XRD intensities of the (002) supercell reflection. The destruction of short-range order was confirmed from the reductions in the Raman intensities and disappearances of Raman bands with fluence. The disordering is more efficient with the increase in the electronic stopping power than with fluence or accumulated electronic energy deposition. These disorders due to the irradiation are discussed in relation to irradiation parameters such as electronic stopping power, accumulated electronic energy deposition and fluence. [Copyright &y& Elsevier]

Abstract: A SiC fiber reinforced SiC composite with magnesium–silicon-based oxide interphase was fabricated by the chemical vapor infiltration process and its mechanical properties were evaluated. An interphase was prepared on the advanced SiC fiber, Tyranno SA, using the alkoxide method. The tensile strengths of SiC fiber with the Mg–Si–O coating retained at 85% or higher compared to uncoated-unheated fiber after heating below the 1300°C, while strength were slightly degraded to 80% after heating at 1400°C. The composite showed ductile failure behavior and the fiber pull-out effect was observed in the fracture surface of the composite. [Copyright &y& Elsevier]

Abstract: The effect of displacement damage up to 50dpa on microstructural development in SiC/SiC composites with Hi-Nicalon Type S SiC fiber was investigated. He bubbles were observed in the matrix of SiC/SiC composites irradiated to more than 10dpa in ion irradiations at 1000°C. The average size of He bubbles increased with increasing displacement damage. Almost all the He bubbles were formed at the grain boundaries in the matrix irradiated to 10dpa. On the other hand, He bubbles were also formed within the grains of the matrix irradiated to 50dpa. The matrix irradiated to 50dpa in a fusion reactor relevant condition had the largest average size of He bubbles. The average size of He bubbles in the matrix irradiated to 50dpa decreased with increasing the amount of implanted H to more than a fusion reactor relevant condition. He bubbles were formed in fibers irradiated to more than 10dpa. [Copyright &y& Elsevier]

Abstract: Fracture toughness specimens of the ferritic–martensitic steel JLF-1 were investigated before and after irradiation at two different temperatures in the High Flux Isotope Reactor. Small (12.5mm in diameter with thickness of 4.6mm) disk-shaped compact tension specimens were irradiated at average temperatures of ∼250°C and 377°C to ∼4dpa. Small, 3.33×3.33×25mm, pre-cracked Charpy specimens were irradiated at ∼300°C and 500°C to 5dpa. Transition fracture toughness was evaluated in terms of the reference temperature T 0 for each irradiation temperature and dose and compared to unirradiated T 0. Current fracture toughness shifts compared with T 0 shifts of F82H and 9Cr2WVTa steels irradiated at similar conditions. The present results show that JLF-1, F82H, and 9Cr–2WVTa steels have very similar resistance to radiation embrittlement after doses of 4–5dpa in the temperature range from 250°C to 500°C. [Copyright &y& Elsevier]

Abstract: Effects of heat treatment and irradiation on mechanical properties and microstructures have been studied for martensitic steel F82H co-doped with 60ppm B and 200ppm N (F82H+B+N) to evaluate fundamental mechanical properties and irradiation response before irradiation at JMTR and HFIR facilities. The specimens were firstly normalized at 1150°C and tempered at 700°C, secondly normalized at 1000°C and tempered at 700, 750 and 780°C. The tensile properties were measured for the specimens before irradiation. Single ion irradiations of 10.5MeV Fe3+ and dual ion irradiations of 10.5MeV Fe3+ with simultaneous 1.05MeV He+ of 10appmHe/dpa rate were performed at 160–590°C to 20dpa. Micro-hardness was measured before and after the irradiation. Tensile properties of the F82H+B+N were similar to F82H and also radiation hardening behaved similarly to F82H. The change of hardening increased with increasing temperature, saturated around 350°C and decreased at higher temperature. [Copyright &y& Elsevier]

Abstract: Reduced-activation martensitic steel F82H was irradiated at alternating temperatures between 230 and 350°C to the accumulated damage level of 1.5dpa using an irradiation capsule with temperature control independent of reactor power. Tensile tests were conducted in order to investigate the effects of the irradiation temperature variations on mechanical properties of F82H. Electron microscope observations were performed for the irradiated F82H to evaluate microstructural evolution of the specimens following varying temperature irradiation. Yield stress of the F82H irradiated at 50% alternating temperature between 230 and 350°C was relatively large compared with the other temperature variations in this study. Size and number densities of dislocation loops were observed to be affected by changing irradiation temperature. The distinctive hardening behavior could be interpreted by the difference in the size and density of the defect clusters in terms of the effect of varying temperature. [Copyright &y& Elsevier]

Abstract: The effect of Al and Be ions pre-implantation on microstructural change and, the formation and growth of He bubbles in SiC/SiC composite was investigated. Four kinds of ion implanted specimens were prepared with 100appm Al, 1000appm Al, 100appm Be and 1000appm Be implanted. No microstructural change was observed in the matrices and fibers of SiC/SiC composites implanted with Al or Be ions up to 1000appm. The un-implanted and Al or Be pre-implanted SiC/SiC composites were simultaneously irradiated to 10dpa using triple ion-beams (6.0-MeV Si2+, 1.0-MeV He+ and 340-keV H+) at 1000°C. Helium bubbles were formed in every matrix and fiber irradiated by triple ion-beams. The size of He bubbles in the matrix was increased by implanting Al or Be ions and increased with increasing amount of implanted Al or Be ions. The size of He bubbles in the fiber was slightly increased by implanting Al or Be ions. These results suggest that Al or Be as transmutation products and impurities may accelerate the growth of He bubbles in SiC/SiC composites under fusion reactor conditions. [Copyright &y& Elsevier]

Abstract: Tensile and Charpy specimens of 9Cr–1MoVNb (modified 9Cr–1Mo) and 12Cr–1MoVW (Sandvik HT9) steels and these steels doped with 2% Ni were irradiated at 300 and 400°C in the High Flux Isotope Reactor (HFIR) up to ≈12dpa and at 393°C in the Fast Flux Test Facility (FFTF) to ≈15dpa. In HFIR, a mixed-spectrum reactor, (n, α) reactions of thermal neutrons with 58Ni produce helium in the steels. Little helium is produced during irradiation in FFTF. After HFIR irradiation, the yield stress of all steels increased, with the largest increases occurring for nickel-doped steels. The ductile–brittle transition temperature (DBTT) increased up to two times and 1.7 times more in steels with 2% Ni than in those without the nickel addition after HFIR irradiation at 300 and 400°C, respectively. Much smaller differences occurred between these steels after irradiation in FFTF. The DBTT increases for steels with 2% Ni after HFIR irradiation were 2–4 times greater than after FFTF irradiation. Results indicated there was hardening due to helium in addition to hardening by displacement damage and irradiation-induced precipitation. [Copyright &y& Elsevier]

Abstract: In part I of this helium-effects study on ferritic/martensitic steels, results were presented on tensile and Charpy impact properties of 9Cr–1MoVNb (modified 9Cr–1Mo) and 12Cr–1MoVW (Sandvik HT9) steels and these steels containing 2% Ni after irradiation in the High Flux Isotope Reactor (HFIR) to 10–12dpa at 300 and 400°C and in the Fast Flux Test Facility (FFTF) to 15dpa at 393°C. The results indicated that helium caused an increment of hardening above irradiation hardening produced in the absence of helium. In addition to helium-effects studies on ferritic/martensitic steels using nickel doping, studies have also been conducted over the years using boron doping, ion implantation, and spallation neutron sources. In these previous investigations, observations of hardening and embrittlement were made that were attributed to helium. In this paper, the new results and those from previous helium-effects studies are reviewed and analyzed. [Copyright &y& Elsevier]

Abstract: Small specimen test technology (SSTT) has been developed to investigate mechanical properties of nuclear materials. SSTT has been driven by limited availability of effective irradiation volumes in test reactors and accelerator-based neutron and charged particle sources. In this study, new bend test machines have been developed to obtain fracture behaviors of F82H steel for very small bend specimens of pre-cracked t/2-1/3CVN (Charpy V-notch) with 20mm length and deformation and fracture mini bend specimen (DFMB) with 9mm length and disk compact tension of 0.18DCT (disk compact tension) type, and fracture behaviors were examined at 20°C. The effect of specimen size on ductile–brittle transition temperature (DBTT) of F82H steel was examined by using 1/2t-CVN, 1/3CVN and t/2-1/3CVN, and it was revealed that DBTT of t/2-1/3CVN and 1/3CVN was lower than that of t/2-CVN. DBTT behaviors due to helium and displacement damage in F82H-std irradiated at about 120°C by 50 or 100MeV He ions to 0.03dpa were also measured by small punch tests. [Copyright &y& Elsevier]

Abstract: The reduced activation ferritic steel (RAFS) has been selected as structural material for a variety of blanket systems for ITER test blanket modules (TBM). In the evaluation of integrated performance of ferritic steels as structural components of blanket systems, there are unique issues as well as common issues for each blanket system. One of the unique issues for each system is the compatibility of ferritic steels with the coolant materials. The corrosion rate of ferritic steels in hot water, super critical pressurized water (SCPW), humid air, Pb–17Li, lithium and Flibe at various temperatures is reviewed in this work. Efforts to improve corrosion resistance have been made, taking the alloy design into account. A dispersion of yttria was effective to improve corrosion resistance of a RAFS. The compatibilities of RAFSs with hot water, Pb–17Li, lithium and Flibe are considered to be good enough for the TBM applications. The liquid metal embrittlement (LME) is considered to be a critical issue for the utilization of RAFSs for the lithium systems. Several issues towards DEMO and beyond are shown from the compatibility point of view. [Copyright &y& Elsevier]

Abstract: Two types of thick (∼3μm) SiC interphase layers with new concept between fiber and matrix were prepared; the porous SiC interphase by polymer impregnation and pyrolysis (PIP) treatment, and the two-ply interphase consisting of carbon and β-SiC by chemical vapor deposition (CVD) treatment. The SiC/SiC composites with these new interphase layers were fabricated by reaction bonding (RB) process. The effect of these interphase layers on microstructure, mechanical and thermal properties of SiC/SiC composites was investigated. The densities of SiC/SiC composites in this study attained to relative densities of 92%. The microstructural observation revealed that the two-ply interphase by CVD treatment prevented the fibers from reacting with the melting Si during RB process. This effect leads to the fiber pull-out phenomenon in the specimen with the two-ply interphase, and therefore this specimen exhibited non-catastrophic failure behavior and high bending strength. The thermal conductivities of specimens in this study were much higher than those of the composites by conventional process. The relative density and thermal conductivity of SiC/SiC composites in this study are high enough to attain the assumed design criteria for fusion reactors. [Copyright &y& Elsevier]

Abstract: The process optimization for the forced-flow/thermal gradient chemical vapor infiltrated SiC based composites with an advanced SiC fiber(Tyranno SA) was carried out. The new SiC/SiC composites had a lower porosity and the uniform distribution of pores compared with conventional CVI. The uniform interphases between SiC fibers and matrix could be obtained by reversing the gas-flow direction mid-way through the coating process. The tensile strength was slightly increased with the thickness of carbon interphase in the range of 20–250nm. It was found that the fabric layer orientation and multilayer SiC/C interphase were very effective to improve the mechanical properties. [Copyright &y& Elsevier]

Abstract: The microstructure of near-stoichiometric fiber SiC/SiC composites implanted with He and H ions was studied at implantation temperatures of 1000 and 1300°C. The average size of He bubbles in the CVI SiC matrix decreases with increasing concentration of implanted H ions. Moreover, the number density of He bubbles increases with increasing irradiation temperature and amount of implanted H. At the irradiation temperature of 1000°C, He bubbles were mainly formed at grain boundary within the matrix. On the other hand, He bubbles were formed both at grain boundaries and within grains at the irradiation temperature of 1300°C. The average size of He bubbles at grain boundaries was much larger than within the grain. The average size of He bubbles in the fiber was smaller than that in the matrix in all cases. [Copyright &y& Elsevier]

In the fusion reactor design study at Japan Atomic Energy Institute (JAERI), several waste management strategies were assessed. The assessed strategies are: (1) reinforced neutron shield to clear the massive ex-shielding components from regulatory control; (2) low aspect ratio tokamak to reduce the total waste; (3) reuse of liquid metal breeding material and neutron shield. Combining these strategies, the weight of disposal waste from a low aspect ratio reactor VECTOR is expected to be comparable with the metal radwaste from a light water reactor (∼4000 t). [Copyright &y& Elsevier]

Micro-indentation hardness was measured for the irradiated F82H steels by single (10.5 MeV Fe3+) beam or dual (10.5 MeV Fe3+ and 1.05 MeV He+ ions) beam at the TIARA facility in JAERI. The extra component of radiation hardening due to helium was slightly detected in the dual-beam (10 appmHe/dpa) irradiation at 633 K up to 33 dpa. As increased the ratio of He/dpa (100 appmHe/dpa), the extra component due to helium was increased. The microstructures in single/dual (10 appmHe/dpa) ion beam irradiated F82H steels consisted of interstitial loops and defect clusters at 50 dpa. However, at a higher ratio of He/dpa (100 appmHe/dpa), nano-voids were also observed in dual ion irradiated F82H. [Copyright &y& Elsevier]

SiC/SiC composite with SiC/C multi-layer interphase coated on advanced SiC fibers was fabricated by the forced thermal-gradient chemical vapor infiltration (F-CVI) process. SEM and TEM observations verified that SiC/C multi-layer interphase was formed on SiC fibers. Both flexural and tensile strengths of SiC/SiC composite with SiC/C multi-layer interphase were approximately 10% higher than composites fabricated with single carbon interphase. The SEM observation of fracture surface for the composite with SiC/C multi-layer interphase revealed cylindrical steps formed around the fiber. Apparently several crack deflections occurred within SiC/C multi-layer interphase. Moreover, the SiC/C multi-layer applied in this study operated efficiently to improve the mechanical properties. [Copyright &y& Elsevier]

New silicon-based oxide layers, SiO2 and SiO2–MgO, as the interfacial materials of SiC/SiC composites were prepared on Hi-Nicalon Type S SiC fiber by sol–gel method. The fibers were completely coated by only dipping twice in a coating solution of [Si]=1.0 mol/dm3 or that of [Mg]=0.50 mol/dm3 and [Si]=0.25 mol/dm3. These coated layers were amorphous up to 1200 °C for the SiO2 coated fibers, or consisted of a mixture of SiO2 and MgxSizO up to 1400 °C in SiO2–MgO coated fibers. The tensile strength of coated Hi-Nicalon Type S SiC fiber after heating at 1200 °C was similar to that of unheated Hi-Nicalon Type S SiC fiber without heating and was reduced by 15% after heating to 1400 °C. [Copyright &y& Elsevier]

The objectives of this work are to evaluate the effects of thermal history corresponding to a blanket fabrication process on Reduced Activation Ferritic/Martensitic steel (RAF/Ms) microstructure, and to establish appropriate Hot Isostatic Pressing (HIP) conditions without degradation in the microstructures. One of RAF/Ms F82H and its modified versions were investigated by metallurgical methods after isochronal heat treatments up to 1473 K simulating HIP thermal history. Although conventional F82H showed significant grain growth after conventional solid HIP conditions, F82H with 0.1 wt% tantalum maintained a fine grain structure after the same heat treatment. It is considered that the grain coarsening was caused by dissolution of tantalum-carbide which immobilizes grain boundaries. On the other hands, conventional RAF/Ms with coarse grains were recovered by post HIP normalizing at temperatures below the TaC solvus temperature. This process can refine the grain size of F82H to more than ASTM grain size number 7. [Copyright &y& Elsevier]

The effects of irradiation on the Charpy impact properties of reduced-activation ferritic/martensitic steels were investigated on a microstructural basis. It was previously reported that the ductile–brittle transition temperature (DBTT) of F82H-IEA and its heat treatment variant increased by about 130 K after irradiation at 573 K up to 5 dpa. Moreover, the shifts in ORNL9Cr–2WVTa and JLF-1 steels were much smaller, and the differences could not be interpreted as an effect of irradiation hardening. The precipitation behavior of the irradiated steels was examined by weight analysis and X-ray diffraction analysis on extraction residues, and SEM/EDS analysis was performed on extraction replica samples and fracture surfaces. These analyses suggested that the difference in the extent of DBTT shift could be explained by (1) smaller irradiation hardening at low test temperatures caused by irradiation-induced lath structure recovery (in JLF-1), and (2) the fracture stress increase caused by the irradiation-induced over-solution of Ta (in ORNL9Cr–2WVTa). [Copyright &y& Elsevier]

Samples of polycrystalline ceramics of MgAl2O4 irradiated at the ambient temperature with 85 MeV I7+ iodine ions to doses up to 1×1019 m−2 is observed to be amorphized up to depths around 6 μm from the ion-incident surface for a dose of 1.2×1019 m−2, through a cross-sectional transmission electron microscopy. A step height of 1 μm is formed across the border between the masked and irradiated regions of the surface. The height of the step is observed to increase sharply from the irradiated area towards the edge at the border, forming a peak as tall as 1.5 μm.A glossy, silver-gray film with a thickness less than 0.1 μm is unexpectedly observed to have formed on the surface of samples of MgAl2O4 and MgO, in about 3.5 years aging after the irradiation to 1.2×1019 m−2, being left untouched in the air. The film is easily peeled off along grain boundaries and found to be amorphous from the electron diffraction pattern. The film from MgAl2O4 sample contains Al, Mg and Si. Silicon, which is one of impurities, is found to be enriched noticeably at the surface. [Copyright &y& Elsevier]

Samples of polycrstalline sintered ceramics of Al2O3, MgAl2O4 and MgO were irradiated at room temperature with 85 MeV I7+ ions to doses up to 1.2×1019 m2. Both MgAl2O4 and Al2O3 samples are observed to be amorphized up to depths of approximately 5–6 μm for a dose of 1.2×1019/m2, through a cross-sectional transmission electron microscopy (XTEM). No defected clusters are observed at approximately 7.5–8.5 μm, where displacement damage due to the nuclear energy deposition is predicted to peak at 1 dpa (displacement per atom). No amorphization is observed for the MgO sample irradiated with the same dose. Instead, X-ray diffractometry (XRD) revealed an enhancement of the diffraction peaks for the (200), and (400) reflections and a reduction of peak intensities from other reflections, as compared with those before irradiation. This indicates that an atomistic rearrangement may occur along an ion path to form a new surface with the lower energy. [Copyright &y& Elsevier]

Microstructure in single crystalline Al2O3 developed during irradiation by swift heavy ions has been investigated. The specimens were irradiated by Xe ions with energies from 70 to 160MeV at ambient temperature. The fluences were in the range from 1.0×1013 to 1.0×1015 ions/cm2. After irradiations, X-ray diffractometry (XRD) measurements and cross sectional transmission electron microscope (TEM) observations were conducted. The XRD results indicate that in the initial stage of amorphization in single crystalline Al2O3, high-density S e causes the formation of new planes and disordering. The new distorted lattice planes formed in the early stage of irradiation around the fluence of 5.0×1013 ions/cm2 for single crystalline Al2O3 irradiated with 160MeV-Xe ions. Energy dependence on structural modification was also examined in single crystalline Al2O3 irradiated by swift heavy ions. The XRD results indicate that the swift heavy ion irradiation causes the lattice expansion and the structural modification leading to amorphization progresses above the energy around 100MeV in this XRD study. The TEM observations demonstrated that amorphization was induced in surface region in single crystalline Al2O3 irradiated by swift heavy ions above the fluence expected from the results of XRD. Obvious boundary was observed in the cross sectional TEM images. The crystal structure of surface region above the boundary was identified to be amorphous and deeper region to be single crystal. The threshold fluence of amorphization was found to be around 1.0×1014 ions/cm2 in the case over 80MeV swift heavy ion irradiation and the fluence did not depend on the crystal structures. [ABSTRACT FROM AUTHOR]

Abstract: Irradiation hardening and fracture toughness of reduced-activation ferritic/martensitic steel F82H after irradiation were investigated with a focus on changing the fracture toughness transition temperature as a result of several heat treatments. The specimens were standard F82H-IEA (IEA), F82H-IEA with several heat treatments (Mod1 series) and a heat of F82H (Mod3) containing 0.1% tantalum. The specimens were irradiated up to 20dpa at 300°C in the High Flux Isotope Reactor under a collaborative research program between JAEA/US-DOE. The results of hardness tests showed that irradiation hardening of IEA was comparable with that of Mod3. However, the fracture toughness-transition temperature of Mod3 was lower than that of IEA. The transition temperature of Mod1 was also lower than that of the IEA heat. These results suggest that optimization of specifications on the heat treatment condition and modification of the minor alloying elements seem to be effective to reduce the fracture toughness-transition temperature after irradiation. [Copyright &y& Elsevier]

Single-phase silicon carbide (SIC) nanotubes were successfully synthesized by the reaction of carbon nanotubes with silicon powder at 1200°C for 100 h. X-ray diffraction patterns indicated that most of the carbon from the carbon nanotubes that were reacted with silicon at 1200°C for 100 h was transformed to SiC. Transmission electron microscopy observations revealed that both single-phase SiC nanotubes and C-SiC coaxial nanotubes, which are carbon nanotubes sheathed with a SiC layer, were synthesized after 100 h of reaction, The ratio of single-phase SiC nanotubes to C-SiC nanotubes increased with heat treatment at 600°C in air for 1 h because the remaining carbon was removed. [ABSTRACT FROM AUTHOR]

Abstract: Several types of reduced activation ferritic/martensitic (RAFM) steel have been developed over the past 30years in China, Europe, India, Japan, Russia and the USA for application in ITER test blanket modules (TBMs) and future fusion DEMO and power reactors. The progress has been particularly important during the past few years with evaluation of mechanical properties of these steels before and after irradiation and in contact with different cooling media. This paper presents recent RAFM steel results obtained in ITER partner countries in relation to different TBM and DEMO options. [Copyright &y& Elsevier]