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AbstractThis paper provides an overview of Japanese fusion engineering research activities focusing on those being carried out by the National Institute for Fusion Science (NIFS) and Japanese universities (Universities). NIFS is promoting the Fusion Engineering Research Project (FERP) as one of three research projects. The majority of the activity in FERP is being carried out by collaboration with Universities. Utilizing the core facilities installed in NIFS and the unique infrastructures of Universities, collaboration between NIFS and Universities is performed for the superconducting magnet, the liquid breeder blanket, advanced materials, high heat flux components, and tritium safety. NIFS also carries out international collaboration programs such as Japan-China–based, Japan-U.S.–based, and International Energy Agency–based collaborations, promoting participation of University researchers. Division of responsibilities with the National Institutes for Quantum and Radiological Science and Technology (QST), contributions to the ITER Broader Approach, and the Action Plan Toward DEMO Development are also reported.

Nd,Y:CaF_2 and Nd,La:CaF_2 ceramics featuring good optical quality have been fabricated by reactive sintering and a hot isostatic pressing method. The transmission spectra, emission spectra, and fluorescence decay curves were measured. Lasing at 1064 and 1065 nm was observed in Nd,Y:CaF_2 and Nd,La:CaF_2, respectively, upon quasi-continuous-wave pumping by a diode laser emitting at 791 nm. To the best of our knowledge, this is the first demonstration of Nd^3+-activated CaF_2 ceramic laser.

Fusion materials research started in the early 1970s following the observation of the degradation of irradiated materials used in the first commercial fission reactors. The technological challenges of fusion energy are intimately linked with the availability of suitable materials capable of reliably withstanding the extremely severe operational conditions of fusion reactors. Although fission and fusion materials exhibit common features, fusion materials research is broader. The harder mono-energetic spectrum associated with the deuterium–tritium fusion neutrons (14.1?MeV compared to <2?MeV on average for fission neutrons) releases significant amounts of hydrogen and helium as transmutation products that might lead to a (at present undetermined) degradation of structural materials after a few years of operation. Overcoming the historical lack of a fusion-relevant neutron source for materials testing is an essential pending step in fusion roadmaps. Structural materials development, together with research on functional materials capable of sustaining unprecedented power densities during plasma operation in a fusion reactor, have been the subject of decades of worldwide research efforts underpinning the present maturity of the fusion materials research programme.

AbstractErbium oxide (Er2O3) coating layer is one of the suitable coatings to realize an advanced breeding blanket system because it has high electrical resistivity and hydrogen permeation suppression effect. In order to enhance these properties of Er2O3coating, it is necessary to form a thick coating layer with high crystallinity. The formation of a double stacked coating layer on an austenitic stainless steel 316 substrate using an intermediate layer (buffer layer) was investigated for the thicker and high crystallinity of Er2O3coating formation. Yttrium oxide (Y2O3) and cerium oxide (CeO2) were selected as buffer layer between the Er2O3layer and austenitic stainless steel 316 substrate due to their similar lattice constant to that of Er2O3crystal. The texture and grain growth direction of Er2O3was controlled by the Y2O3and CeO2buffer layer. However, the suppression effect of hydrogen permeation by the double stacked coating was smaller than that of the single layer coating due to the thin Er2O3formation.

AbstractPrevious studies by the authors showed that hydrogen permeation reduction factor (PRF) of Er2O3coating on ferritic steels by Metal Organic Decomposition (MOD) depends on the Cr level of the substrate steels and the annealing conditions. The reason of the dependence was attributed to the composition of the oxide layer formed beneath the coating. The PRF was shown to be larger when Cr2O3layer was formed than when Fe2O3layer was formed. This paper reports further investigation of the effect of temperature and oxygen partial pressure of the annealing on the composition of the oxide layer. A diffusion modeling of Cr and O was performed to account for the experimental data. The results showed that the data can be well explained assuming that Cr2O3layer is formed when supply of Cr exceeds that of O at the substrate surface.

AbstractMetal Organic Chemical Vapor Deposition (MOCVD) process is a vapor phase growth technique which is synthesized via vapor phase from metal organic complex material and will become one of the desirable deposition methods to form oxide layer on the complicated shape ducts and large-area walls of breeding blanket components. The single phase erbium oxide (Er2O3) coating layers were able to be synthesized on metal substrates such as stainless-steel 316 (SUS316) and vanadium alloy and Si single crystal plates. In this paper, the hydrogen permeation and electrical property of Er2O3coating layer via MOCVD process will be mainly reported. The hydrogen permeation quantity of Er2O3coating layer estimated by QMS was decreased by 1/20 compared with SUS 316 at 400 °C. In addition, the demonstration of Er2O3coating into the interior surface of the stainless-steel was also succeeded.

AbstractTo examine the accuracy in the neutronics calculations for the Li/V-alloy blanket system without Be neutron multiplier, a fusion neutronics experiment on a Li/V-alloy assembly has been performed with a 14 MeV neutron source. Reaction rates and tritium production rates (TPRs) in the assembly were measured with activation foils and Li2CO3pellets. The measured reaction rates sensitive to fast neutrons agreed almost within ~10 % with ones calculated by using the MCNP5 code, JENDL-3.3 library and JENDL dosimetry file 99. Though there appeared a possibility of a significant underestimation in the transport calculations for the energy range of <~4 keV due to nuclear data of vanadium, the measured TPR was consistent with the calculated one within ~8 %.

AbstractThe hardness distribution and the effect of post welding heat treatment (PWHT) at 600°C and 1000°C for 1 h and aging heat treatment at 600°C for 1000 h in the dissimilar-material electron beam weld (EBW) joint with pure-V and SUS316L austenitic stainless steel were investigated. The electron beam was positioned just on the butt joint (EB00), shifted by 0.2 mm (EB02) and 0.4 mm (EB04) on the pure-V side. The EBW joint was distinguished into the base metal of V (V-BM), weld metal (WM), interlayer at the edge of the WM of SUS316L side (IL) and base metal of SUS316L (SUS316L-BM). The IL was observed only in the EB02 and EB04 joints. The formation of macro-pore was observed in the EB04 joint. Much higher hardness was observed at the WM and IL of the as-welded EB00 and EB02 joints than the other regions of them. The hardness change in the WM was relatively small due to the PWHT at 600°C up to 1 h, whereas significant increment was observed due to the PWHT at 1000°C for 1 h regardless of the EB position. The hardness of the IL after the PWHT at 600°C was almost twice higher than that of the as-welded one, which showed slightly further increment at 1000°C. Rapid increment of the hardness due to the aging at 600°C for 1–10 h and slightly further increment of it due to the aging for 100–500 h occurred in the WM of the EB00 joint.

AbstractIn this work, corrosion experiments on 9Cr-ODS and CLAM steels were carried out in static Pb-Li at 873 K for 250 h. Both steels showed weight loss and softening near the surface after the exposure. Tensile properties did not change and creep properties degraded slightly for 9Cr-ODS steel. In contrast, CLAM steel showed hardening by increase in tensile strength and creep rupture time, and decrease in minimum creep rate and reduction of area. The metallurgical analyses showed that the both steels were non-uniformly corroded by preferential corrosion at grain and sub-grain boundaries. Near the surface, carbides were lost and Cr was depleted to several tens of μm depth. The depletion was heavier for 9Cr-ODS than for CLAM. The corrosion mechanism was proposed to be a loss of protective oxide layer followed by dissolution of Cr in matrix into liquid Pb-Li. The more pronounced corrosion effect on 9Cr-ODS than on CLAM may be due to finer grain and sub-grain size enhancing preferential attack by Pb-Li at the boundaries, or lack of Mn in 9Cr-ODS, which can form protective layers for CLAM.

AbstractThe effects of post-welding heat treatment (PWHT) at temperatures ranging from 640°C to 750°C for 1 h on the mechanical and metallographical properties of a dissimilar-metal electron beam weld (EBW) joint of reduced activation ferritic/martensitic steel, F82H, and SUS316L austenitic stainless steel were investigated. The EBW joint is demarcated into five regions: the base metal of F82H (F82H-BM), the heat-affected zone of F82H (F82H-HAZ), the interlayer at the edge of F82H-HAZ (IL), the weld metal (WM), and the base metal of SUS316L (SUS316L-BM). No hardening resulting from welding and no significant change in the hardness resulting from PWHT above 640°C were observed in the F82H-BM, SUS316L-BM and WM. However, a significantly higher hardness was observed in the as-welded F82H-HAZ and IL than in other regions, and a significant reduction in hardness occurred in F82H-HAZ and IL, as a result of PWHT above 640°C. Irradiation hardening after PWHT at 720°C was investigated using proton-irradiation at 300°C up to 0.1 and 1 dpa. The irradiation hardening of the WM, IL and SUS316L-BM, which was less than that of the as-received SUS316L, was much larger than that of F82H-HAZ and F82H-BM.

AbstractJapan-US cooperation program TITAN (Tritium, Irradiation and Thermofluid for America and Nippon) started in April 2007 as 6-year project. This is the summary report at the midterm of the project. Historical overview of the Japan-US cooperation programs and direction of the TITAN project in its second half are presented in addition to the technical highlights.

AbstractCombination of liquid lithium with Reduced Activation Feritic/Martensitic Steel (RAFM) is one of the options for Test Blanket Module (TBM) in early ITER period and early DEMO blanket, as well as an intermediate step toward Li/V DEMO blanket. In this paper, characterization of a Li/RAFM blanket was carried out and compared with a Li/V blanket from neutronics and compatibility viewpoints.Although the local Tritium Breeding Ratio (TBR) will be reduced by ∼0.1 by the change from Li/V to Li/RAFM, Li/RAFM seems to be still feasible, with enhanced neutron shield, from the tritium self-sufficiency viewpoint. A similar tritium production rate for the Li/V and the Li/RAFM TBMs suggests that the Li/RAFM TBM simulates well the tritium production of Li/V TBM and thus will be suitable for Li/V DEMO blanket design as well as Li/RAFM blanket.Based on the available data, V-alloys are thought to be highly compatible with Li when the impurity level in the Li is low. New compatibility experiments of RAFM with Li showed transformation of martensitic to ferritic phase in addition to corrosion loss. However, the compatibility issue is estimated to be small for ITER-TBM conditions.The present study showed the significance of starting with a Li/RAFM TBM during the early phase of ITER operation for development of both Li/RAFM and Li/V DEMO blankets.

Dissimilar metal electron beam welding with reduced activation ferritic/martensitic steel, F82H IEA heat, and SUS316L austenitic stainless steel was studied. Mechanical property evaluation at room temperature by bend test, tensile test, Vickers hardness measurement and charpy impact test, and evaluation of irradiation hardening by proton irradiation at 300°C up to 0.5 dpa were carried out. The mechanical properties of the dissimilar weld were improved by the optimization of the electron beam position in the welding (shifted 0.2 mm on 316L side) and the post-weld heat treatment (PWHT) (750°C x 1 hour). The improvement of the mechanical properties might be due to the fact that the weld metal consisted of the austenitic phase. Smaller irradiation hardening than 316L was observed in the weld metal of the F82H/316L dissimilar weld after PWHT at 750°C for 1 hour, where the electron beam was shifted 0.2 mm on 316L side, though the formation of voids and dislocation loops occurred in the grain matrix of the weld metal.

AbstractCharacteristics of the tritium diffusion coefficient DTin V-4Cr-4Ti alloy, including a bending in the Arrhenius plot of DT, are examined. Based on a trap model, the possible trap sources and their binding energies for tritium in the alloy are evaluated using the experimental data of DTin pure V, which are measured with a tritium tracer method, and the literature data of protium diffusion in V-Ti and V-Cr alloys. The result of the evaluation suggests the presence of two trap sources in the alloy. The first would be attributed to a trap at each substitutional alloying atom which is likely to be Ti. The binding energy EBof 0.08 eV gives the best fit to the observed value of DTabove 300 K. The bending in the Arrhenius plot below 300 K is caused by a second trap site with a higher EB, and a lower concentration than those of each alloying atom. The trap is probably formed by the alloying atoms presence to neighboring Ti atoms. The contribution of Cr atom to the trap effect seems to be rather small in this alloy.

AbstractA series of irradiation experiments were performed using Fusion Neutronics Source (FNS) at JAEA for the investigation of activation of materials relevant to Lithium/vanadium-alloy and Flibe/vanadium-alloy blankets. The specimens of V-4Cr-4Ti, Er and Teflon in 10 mm×10 mm×0.03-0.1 mm were prepared for studying the activation of V-alloy structure, MHD coating of Er2O3, and F in molten salt Flibe, respectively. Be, Li and Li/Be mock-ups were assembled with Be and solid Li blocks in addition to the assembly for direct D-T neutron irradiation to examine the dependence of the activation on neutron spectrum. The neutron spectra in all irradiation locations were calculated using MCNP code and JENDL-3.3 file. The activities of the specimens induced in various neutron fields were measured with a high purity Ge detector (HP-Ge). Experimental analyses were carried out using FISPACT-2001 codes with both EAF-2001 file and FENDL/A-2.0 & FENDL/D-2.0 libraries. The typical calculation/experiment (C/E) values lay in the range of 0.8-1.2. Coarse group treat for (n,σ) reactions, especially in resonance range, could result in overestimation. Use of continuous cross section improves consistency of the calculation with the experiment. However, accurate estimate of spectra is necessary when the flux changes largely with energy around the threshold or the resonance peak.

AbstractOne of the critical issues of Flibe/V-alloy blanket with REDOX control by Be is a large tritium inventory in V-alloy structures. Among the possible solutions to this issue would be to control REDOX not by Be but by addition of MoF6or WF6enhancing the reaction from T2to TF. The present study investigated feasibility of this procedure by thermodynamic and neutronics calculations. Using the blanket dimensions of Force Free Helical Reactor (FFHR), tritium inventory in V-alloy structure and Flibe were estimated based on the calculated equilibrium partial pressures of T2and TF in various cases of REDOX control by MoF6or WF6. Also carried out were neutronics examinations for the impact of Mo or W doping in the blanket. The results showed that the tritium inventory in the blanket area would be less than 100g at the TF level of 0.1 and 1 ppm in Flibe with addition of WF6and MoF6, respectively. WF6doping is far more advantageous than MoF6doping for low activation purposes.

AbstractFe-Cr-W based low activation ferritic steels are widely regarded as promising blanket structural materials for fusion reactors, while liquid lithium breeder/coolant concept provides an attractive option for high efficiency and simplicity of blanket system. However, past compatibility tests of ferritic steels with liquid lithium were almost limited to conventional Fe-Cr-Mo steels. In this study, the corrosion behavior of the candidate reduced activation ferritic steel, JLF-1(Fe-9Cr-2W-0.1C) in lithium was investigated. Static immersion tests were carried out using coupon specimens (16 × 4 × 0.25 mm) at 873K and 973K for 100hr. At 973K, the phase transformation from martensite to ferrite resulted in decrease in hardness from 250 to 140Hv. This seemed to be caused by depletion of C. Examinations of binary Fe-Cr and pure iron were also carried out for comparison with JLF-1.

AbstractTo improve the data accuracy of the neutron emission spectra of the natLi(d,xn) reaction that will be used as the neutron source in the International Fusion Materials Irradiation Facility, the authors have measured the neutron emission spectra from thick and thin lithium targets bombarded by 40-MeV deuterons at the AVF cyclotron (K= 110) facility of Tohoku University. The neutron spectra were measured at nine laboratory angles between 0 and 110 deg with the time-of-flight method using a beam-swinger system and a well-collimated neutron flight channel, obtained over almost the entire energy range of secondary neutrons using a two-gain method. The data obtained revealed the shape and angular dependence of emission spectra up to the high-energy region called a “high-energy tail.” The experimental results are compared with other experimental data and calculations.

We prepared GdBa₂Cu₃O_7−δ (Gd-123) coated conductors by the pulsed laser deposition (PLD) method on PLD-CeO₂/ion-beam assisted deposition (IBAD)-Gd₂Zr₂O₇ (GZO)/hastelloy metal substrate tapes. The Gd-123 film showed a higher critical current density (J_c) in magnetic fields (B) and higher critical temperature (T_c) than those of YBa₂Cu₃O_7−δ (Y-123) film. The Gd-123 film exhibited a high J_c value of 0.19MAcm⁻² at 3T (_c value was 93.8K. The J_c value at 3T was twice as high as that of Y-123 film. Moreover, the angular dependence in magnetic fields of J_c for Gd-123 was also superior to that of Y-123. The reduction of J_c by the magnetic field angle for Gd-123 was less sensitive than that in Y-123. Furthermore, Gd-123 clearly exhibited a peak at 0 (_c on the magnetic field angle curve (J_c–θ), while Y-123 had no peak at 0. These superior characteristics are considered to be due to the existing defects, such as stacking faults, as observed by transmission electron microscopy. PLD-Gd-123 coated conductors, with pinning-effective defects, high J_c and low anisotropy in J_c–θ, show promise for future applications.

We prepared GdBa2Cu3O7??(Gd-123) coated conductors by the pulsed laser deposition (PLD) method on PLD-CeO2/ion-beam assisted deposition (IBAD)-Gd2Zr2O7(GZO)/hastelloy metal substrate tapes. The Gd-123 film showed a higher critical current density (Jc) in magnetic fields (B) and higher critical temperature (Tc) than those of YBa2Cu3O7??(Y-123) film. The Gd-123 film exhibited a high Jcvalue of 0.19 MA cm?2at 3 T () in liquid nitrogen (77 K), and the Tcvalue was 93.8 K. The Jcvalue at 3 T was twice as high as that of Y-123 film. Moreover, the angular dependence in magnetic fields of Jcfor Gd-123 was also superior to that of Y-123. The reduction of Jcby the magnetic field angle for Gd-123 was less sensitive than that in Y-123. Furthermore, Gd-123 clearly exhibited a peak at 0° () as well as at 90° () in the relationship of Jcon the magnetic field angle curve (Jc-?), while Y-123 had no peak at 0°. These superior characteristics are considered to be due to the existing defects, such as stacking faults, as observed by transmission electron microscopy. PLD-Gd-123 coated conductors, with pinning-effective defects, high Jcand low anisotropy in Jc-?, show promise for future applications.

AbstractJapanese activities on fusion structural materials R & D have been well organized under the coordination of university programs and JAERI/NIMS programs more than two decades. Where, two categories of structural materials have been studied, those are; reduced activation martensitic/ferritic steels (RAFs) as reference material and vanadium alloys and SiC/SiC composite materials as advanced materials. The R & D histories of these candidate materials and the present status in Japan are reviewed with the emphasis on materials behavior under radiation damage. The on-going Japan-USA collaborations on blanket engineering and materials, JUPITER-II Program and on neutron radiation effects, JAERI/ORNL Phase IV and V are also presented.

AbstractA neutronics-based investigation was carried out for self-cooled Blanket Test Modules in ITER with coolant of liquid lithium and structural materials of V-4Cr-4Ti. The design feasibility was demonstrated for the test module having the purpose to validate the tritium self sufficiency for DEMO Li/V blanket without neutron multiplier Be.

The Nagoya Coated Conductor Centre (NCCC) in SRL has installed new reel-to-reel IBAD and PLD systems for the fabrication of long, coated conductors. A 53 m long ion-beam assisted deposition (IBAD) substrate has been successfully produced. The degree of bi-axial texturing was improved by the newly found buffer deposition technique of pulsed laser deposition- (PLD)-CeO₂, which we call the self-epitaxial method. The PLD-CeO₂ layer brought about a high degree of in-plane texturing up to 2.6° in a few minutes on an IBAD seed layer. The resultant YBCO coated conductor on the CeO₂ layer showed high J_c values of up to 4.4 MA cm⁻². The reel-to-reel PLD equipment with a high power laser of 200 W is now being applied to long-conductor fabrication and CeO₂ buffer layer deposition with a high deposition rate.

Research results on corrosion behavior of Flibe (LiF-BeF2molten salt) against structural materials are reviewed. Thermodynamic calculation and static experiment were carried out. It was found that some kinds of oxide films may work as a protective scale for ferritic steel. For V-based alloy, on the other hand, no stable oxide film can exist in the presence of Flibe. It was also found that Redox control might protect V-based alloy against corrosion.

To understand the influence of a stepwise change of irradiation temperature on pure vanadium and vanadium alloys, ion irradiation were performed at 473 K to 0.25 dpa and 873 K to 0.5 dpa. In pure vanadium and V–5Cr alloy, shrinkage of interstitial type loops was prominent during irradiation at 873 K. These results were explained by the appearance of a vacancy rich condition during the successive irradiation at 873 K. In Ti-containing alloys (V–5Ti, V–4Cr–4Ti), on the other hand, dissociation of vacancy loops leads to the shrinkage of interstitial type loops and the formation of Ti-enriched precipitates.

Vanadium alloys are widely regarded as possessing desirable mechanical and physical properties for application as structural materials in fusion power systems. The bulk of the recent research on vanadium is focussed on ternaries containing 4–5% Cr and 4–10% Ti. The aim of this paper is to review significant results generated by the international research and development community on this alloy system and to highlight the critical issues that must be resolved before alloy development can proceed to the next stage. Recent progress on understanding the physical metallurgy, fabrication and joining behavior, and compatibility with hydrogen and oxygen containing environments of unirradiated vanadium alloys is discussed. The effect of low-temperature neutron irradiation on mechanical properties and their relationship to the observed microstructure are briefly summarized. Current efforts to characterize the high-temperature mechanical properties, develop constitutive equations describing flow and fracture, and understand and mitigate the effects of non-metallic impurities on properties are presented.

The influence of temperature variations on microstructural evolution in austenitic stainless steels is discussed in order to help interpret the response of materials in the HFIR-RB-13J temperature variation experiment. A kinetic microstructural evolution model developed for irradiated austenitic stainless steels was modified to provide a fully dynamic calculation of point defect and point defect cluster concentrations. Using the modified model, microstructural evolution was predicted for simulated HFIR-RB-13J temperature variation experiments and variations in material parameters were evaluated. The results indicate that repeated temperature excursion to 573 from 773 K always results in increased dislocation loop density and reduced cumulative defect flux within the calculated material parameter range, while excursions to 473 from 623 K may increase or decrease them depending on material parameters. The primary influence of temperature variation could be explained by accumulation and release of matrix defects at the temperatures of interest. The applicability of the current model must be further studied by careful investigation of the results from the HFIR-13J and related experiments.

The importance of interactions between fusion materials research and development (R&D) and other technologies is emphasized to make attractive and realistic fusion technology integration activities. The focuses are on: (1) materials design and processing, (2) safety issues relating to materials and (3) material performance evaluation methodologies, including 14 MeV neutron source utilization for fusion material R&D. As typical examples, material design activities on reduced activation ferritic steels, vanadium alloys and SiC/SiC composite materials are provided. The safety assessment of reactor systems and reactor design code consideration including prediction methodologies of materials performance are also discussed.

Pure tantalum specimens were irradiated with 2.4 MeV Cu2+ions up to 3 dpa at temperatures between room temperature and 1073 K. Transmission electron microscope (TEM) observation and micro-indentation tests were carried out to correlate the microstructure and the hardness. Significant radiation hardening occurred at temperature ranging from 673 to 873 K. Isochronal annealing of a specimen irradiated at room temperature up to a dose of 0.3 dpa resulted in a rapid increase in hardening between 573 and 673 K and continued to increase up to 873 K. The microstructure showed that the formation of small defect clusters is the major reason for both the radiation hardening and the radiation-anneal hardening.

The National Institute for Fusion Science (NIFS), in collaboration with Japanese industry, has initiated a research program for vanadium alloys, which will be highlighted by the production of a large V–4Cr–4Ti ingot. In this program, the technology and underlying science associated with vanadium alloy fabrication will be enhanced. Effort has been focused on the control of interstitial impurities such as C, O and N. Purification of the present commercial metal vanadium was carried out by improvement of the processes. Medium size (30 kg), high purity V–4Cr–4Ti ingots have been produced, designated as NIFS-HEAT-1, and a large size (200 kg) ingot is planned. In this study, the technology for fabricating large V–Cr–Ti ingots with ∼50 ppm C, ∼180 ppm O and ∼100 ppm N has been demonstrated.

To evaluate the influence of temperature change on microstructural evolution in phosphorus modified Fe–Cr–Ni alloys, irradiations at 473/923, 673/923, 773/923 K were performed on a Fe–16Cr–17Ni–0.1P alloy with 2.4 MeV Cu ions. In these experiments, specimen temperature was changed in a stepwise manner during the irradiation. In the case of 673/923 K-irradiation, the phosphide density after the second irradiation at 923 K was about 2 orders of magnitude higher than that in the constant temperature irradiation at 923 K. The influence of the first irradiation at lower temperature remained to a higher dose level of 75 dpa. High void swelling of about 12% was observed after varying temperature irradiation. In the constant temperature irradiation, on the other hand, void swelling was strongly suppressed. The influence of the first irradiation was relatively small in 473/923 K-irradiation. Defect clusters formed during the first irradiation at 473 K disappeared during temperature elevation up to 923 K, and the microstructural evolution during the second irradiation at 923 K was similar to that in the constant temperature irradiation.

To investigate fatigue properties of a low activation ferritic steel (9Cr–2W steel, JLF-1), low cycle fatigue tests were performed in air at room temperature under axial strain control for a complete push–pull condition. The strain rate was 0.4% s−1. Cyclic strain-hardening was observed within the initial 20 cycles, and then cyclic strain-softening occurred gradually until the final failure, though the plastic strain range did not change significantly. Tensile peak stresses in hysteresis curves measured at around half the number of cycles to failure depended on the total strain range. The drop in the peak stress by the cyclic strain-softening increased with decreasing total strain range. The regression curve of the total strain range against the fatigue life was formulated using the Manson–Coffin equation and the fatigue life of JLF-1 steel was compared with that of 8Cr–2W steel.

The growth of Ni3Si surface films on Ni-12.7at%Si alloys has been measured during lMeV electron irradiation. Stereoscopic techniques were used to determine film thickness from dark field images formed from Ni3Si superlattice reflections. Parabolic growth kinetics are observed at lower temperatures. However, at higher temperatures, deviations from parabolic kinetics are observed after short irradiation times. Such deviations have not been observed in bulk specimens during bombardment with energetic ions and, therefore, may be due to foil thickness effects.

Radiation damage of plasma facing materials and the behavior of impurities in the scrape-off layer, under steady-state discharge of superconducting tokamak TRIAM-1M, was examined with using a rotatable collector probe. TEM observations of metallic probe specimens exposed to the discharges showed that numerous radiation induced defects such as dislocation loops and hydrogen clusters were formed. These are typical defects formed by low energy (< a few keV) hydrogen bombardment. Mo, Fe and Cr, which originated from Mo fixed limiters and the 304SS vacuum vessel walls, were deposited on the specimens. The deposition profile depended on the geometrical relationship with the main plasma and affected the internal damage of the specimens. It is worth noting that the density of the metallic impurities in the scrape-off layer was almost constant during the steady-state discharge and had no adverse effects on plasma confinement in the TRIAM-IM.