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1.2C–4Cr–4Mo–10W–3.5V–10Co–Fe high-speed steel (JIS SKH57; ISO HS10-4-3-10) is often manufactured via casting and forging. By applying powder metallurgy, the properties of the abovementioned material can be improved. In this study, the effects of sintering conditions on the formation of precipitates and pores are evaluated. Additionally, strength with and without hydrostatic pressure during sintering is evaluated via static bending and impact tests. Sintering via hot isostatic pressing (HIP) at 1463 K can effectively eliminate pores and prevent the coarsening of precipitates. Toughness and strength improved by 50% by applying HIP.

The mechanisms of irradiation embrittlement of two Japanese RAFSs were different from each other. The larger DBTT shift observed in F82H is interpreted by means of both hardening effects and a reduction of cleavage fracture stress by M23C6 carbides precipitation along lath block and packet boundaries, while that of JLF-1 is due to only the hardening effect. Dimensional change measurement during in-pile creep tests revealed the creep strain of F82H was limited at 300 °C. Performance of the weld bond under neutron irradiation will be critical to determine the life time of blanket structural components. Application of the ODS steels, which are resistant to corrosion in supercritical pressurized water, to the water-cooled blanket is essential to increase thermal efficiency of the blanket systems beyond DEMO. The coupling of RAFS and ODS steel could be effective to realize a highly efficient fusion blanket.

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 10–12 dpa at 300 and 400 °C and in the Fast Flux Test Facility (FFTF) to 15 dpa 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.

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 ≈12 dpa and at 393 °C in the Fast Flux Test Facility (FFTF) to ≈15 dpa. In HFIR, a mixed-spectrum reactor, ( n , α ) reactions of thermal neutrons with 58 Ni 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.

Among the various tritium transport processes in lithium ceramics, the importance and the detailed mechanism of surface reactions still remains to be elucidated. The dynamic adsorption and desorption (DAD) model has been developed to calculate the tritium release behavior from Li2O surface. In the DAD model, the tritium release behavior is considered to be controlled by the surface coverage of adsorbed species, such as OH−, O2−, H−, determined by the H2 and H2O concentrations in the gas phase and the temperature. In a previous paper, the tritium residence time on the Li2O surface was calculated from the model, and the calculation results were compared with the experimental results. It was shown that the calculation results agreed well with the experimental results. In the present paper, a model for the surface behavior of tritium release from Li2TiO2 was constructed, according to the concept of the model for Li2O. The calculation results are compared with the experimental results and the validity of the tritium release mechanism assumed in the model is discussed.

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.

The tritium release behavior from Li2O sintered pellets (81–88% T.D.) is examined by isothermal heating tests. (1) For the 88% T.D. specimens, the fraction of residual tritium is found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (2) For the 81% T.D. specimens, which are annealed after irradiation at 630 K for 4 h, the fraction of residual tritium is also found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (3) For the 81% T.D. specimens as irradiated, the tritium release rate is found to follow the square-root law of the annealing time. The rate-determining process is controlled by Kohlrauch stretched exponential form. Tritium trapped in irradiation defects released with recovering the defects by isothermal heating.

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 20 mm length and deformation and fracture mini bend specimen (DFMB) with 9 mm 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 100 MeV He ions to 0.03 dpa were also measured by small punch tests.

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 30 appm He.

Preparation conditions of single-phase SiC nanotubes and C-SiC coaxial nanotubes were investigated. The characterization of single-phase SiC nanotubes and C-SiC coaxial nanotubes were carried out. The SiC nanowires, which were made of the catenated SiC grains of 50–200 nm in diameter, were obtained in carbon nanotubes reacted at 1450 °C. The only C-SiC coaxial nanotubes were formed at 1300 °C. A few single-phase SiC nantoubes were synthesized at 1200 °C for 100 h. More than half number of nanotubes reacted at 1200 °C for 100 h were altered to single-phase SiC nantoubes by heat treatment of 600 °C for 1 h in air since the remained carbon was removed. The energy dispersive X-ray spectroscopy analysis revealed that the atomic ratio of Si to C in single-phase SiC nanotubes was almost 1; these single-phase SiC nanotubes consisted of near-stoichiometric SiC grains.

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 2.3 dpa. In this study, 10 B and 11 B-doped specimens were irradiated to minimize the errors from the effect of B on mechanical properties by comparing the results. The specimens used were 10 B-doped, 10 B + 11 B-doped and 11 B-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 330 appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50 MeV-He 2+ irradiation was also performed to implant about 85 appm He atoms at 120 °C by AVF cyclotron to 0.03 dpa, 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 100 MPa m 1/2 DBTT for the F82H + 11 B, F82H + 10 B + 11 B, and F82H + 10 B specimens were 40, 110, and 155 °C, respectively. The shifts of DBTT due to helium production were evaluated as about 70 °C by 190 appm He and 115 °C by 330 appm 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.

Japanese 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 ...

Two types of thick (∼3 μm) 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.

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 20–250 nm. It was found that the fabric layer orientation and multilayer SiC/C interphase were very effective to improve the mechanical properties.

Effects of tempering treatment on mechanical properties and microstructures have been studied for martensitic steel F82H doped with 60 ppm B and 200 ppm N (F82H þ B þ N). The tempering treatments were performed at 700–780 � C after the normalizing treatment at 1000 � C. Yield stress of the F82H þ B þ N steel tempered at 700, 750 and 780 � C was 740, 580 and 500 MPa, respectively, and ductile-brittle transition temperature (DBTT) of the specimens was � 55, � 85 and � 85 � C, respectively. The areal density of dislocations decreased from 1:1 � 10 14 to 2:5 � 10 13 m � 2 with increasing tempering temperature from 700 to 780 � C. The number density of precipitates decreased with increasing tempering temperature from 700 to 750 � C, while the number density was almost equivalent as increasing tempering temperature from 750 to 780 � C. The results indicate that the change of DBTT, depending on tempering temperature, is related with the change of yield strength, size and number density of carbides. Hardening behavior of the F82H þ B þ N steel irradiated by 10.5 MeV Fe 3þ to 10 dpa at 360 � C has been also studied by using a micro-indentator. The micro-hardness of the F82H þ B þ N steel tempered at 780 � C was changed from 3.6 to 4.8 GPa by the irradiation. Because hardening behavior of the F82H þ B þ N steel was found to be similar with that of F82H non-doped, doping effects of B on irradiation hardening were suppressed by co-doping of B and N.

The inftucence of addition of 1 0 B and 1 1 B on radiation hardening as a function of irradiation temperature was examined in reduced-activation martensitic F82H+B steels (8Cr-2W-0.2V-0.04Ta-0.1C-0.006B). Specimens used were 1 0 B doped, 1 0 B+ 1 1 B doped and 1 1 B doped F82H steels. Helium concentration produced in the specimens through 1 0 B(n, α) 7 Li reaction was measured from about 15 to about 330appm. Radiation hardening of the specimens irradiated at 300°C to 2.3 dpa and 150°C to 1.9 dpa was observed. Increment of radiation hardening possibly due to helium in the specimen irradiated at 300°C tended to increase slightly with increasing helium production in tensile test at 20°C, but it was not observed for 150°C irradiation. Therefore, it can be mentioned that the enhancement of radiation hardening in the 1 0 B( 1 1 B) doped steels depends on irradiation temperature, The dependence of radiation hardening on tempering time was also examined for F82H-std irradiated at 150'C to 1.9 dpa. Increment of yield strength of F82H-std tended to increase with increasing tempering time. The relation between the shift of DBTT and the increment of yield strength due to irradiation suggest that the extension of tempering time or the increase of tempering temperature would be beneficial for mechanical properties of F82H-std.

F82H (Fe–8Cr–2W) and its variant doped with 2%Ni were irradiated up to 20 dpa 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.

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.

The structure of (1− x )B 2 O 3 – x PbO ( x =0.20, 0.30, 0.40) glasses and (0.98− x )B 2 O 3 – x PbO–0.02PbBr 2 ( x =0.18, 0.28, 0.38) glasses was studied by time-of-flight pulsed neutron diffraction. The structural parameters for each atomic pair were optimized in the Q -space, and the distances of the near neighbor B–O correlations forming trigonal-plane BO 3 units and tetrahedral BO 4 units in boroxol-like rings were estimated at 0.137 and 0.148 nm, respectively. The existence of the species was also supported by Raman spectroscopy and ab initio molecular orbital calculation. The Pb atoms served as modifiers to hold the networks, the neighboring oxygens among which formed distorted octahedra. The distances of Pb–O pairs were divided into two groups of approximately 0.25 and 0.30 nm, respectively. The medium-range order derived from Pb atoms was recognized in the structure factor S ( Q ) over the composition range of this work. The short-range order in the boron–oxygen network was unaffected by PbBr 2 doping, but Br − ions were coordinated in the oxygen sites around Pb atoms.

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.

Fatigue properties were examined on a reduced activation ferritic/martensitic steel (F82H), and preliminary results are presented. F82H steel was irradiated at 523 K to 3.8 dpa, and then fatigue-tested at 298–573 K in vacuum with total strain range of 0.4–1.0%. The fatigue life of the irradiated specimen tested at 298 K with total strain range of 0.4% was revealed to be reduced to about 1/7 of that for the unirradiated specimen. The reduction of the fatigue life was attributed to the change of the fatigue mechanism to the channel fracture. The effect of test temperature is also discussed.

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).

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.

The irradiation behavior of Li 2 TiO 3 under a fusion reactor environment was simulated by simultaneous irradiation of Li 2 TiO 3 by the triple ion beams and the respective single ion beams of O 2+ , He + and H + . The microstructural changes in Li 2 TiO 3 caused by the irradiation were measured by Raman spectroscopy and FT-IR photoacoustic spectroscopy. The results suggest that the formation of TiO 2 due to displacements by irradiation occurs, and the irradiation defects generated by irradiation trap hydrogen and increase the amount of hydroxyl near the surface. Such phenomena are believed to significantly affect the chemical form of the released tritium and the tritium inventory in the breeding materials of a fusion reactor.

Tritium release rate from Li 2 O (71–86% T.D.) is found to follow the stretched exponential form, d F /d t =exp(−( t / τ )) β . The values of β are about 0.8 near 500 K and about 0.5 near 580 K. The activation energy of tritium release is calculated as approximately 92 kJ/mol at 493–533 K and 139 kJ/mol at 543–583 K. It is suggested that the rate controlling process of tritium release is detrapping from the irradiation defects that serve as trapping sites for tritium, and the recovery behavior of such irradiation defects significantly affects the tritium release behavior. No porosity dependence of tritium release was observed for these densities of Li 2 O.

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).

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.

Micro-indentation hardness was measured for the irradiated F82H steels by single (10.5 MeV Fe 3+ ) beam or dual (10.5 MeV Fe 3+ 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.

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.

New silicon-based oxide layers, SiO 2 and SiO 2 –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/dm 3 or that of [Mg]=0.50 mol/dm 3 and [Si]=0.25 mol/dm 3 . These coated layers were amorphous up to 1200 °C for the SiO 2 coated fibers, or consisted of a mixture of SiO 2 and Mg x Si z O up to 1400 °C in SiO 2 –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.

A small specimen test technique is required to evaluate the fracture toughness values of several millimeter thick plates of structural materials and to maximize the use of very limited space for materials irradiation in intense neutron sources like IFMIF. In view of several advantages of three-point bending (3PB) over compact tension (CT), miniaturized 3PB specimens with 7.0, 5.0 and 3.3 mm thickness were prepared from a Japanese low activation ferritic steel, JLF-1, which is a candidate first wall and fusion blanket material.Elastic-plastic fracture toughness tests by the unloading compliance method at room temperature and plane-strain fracture toughness tests at 77 K were conducted in general accordance with the ASTM standards. Emphasis was focused on the determination of the actual J-value for crack initiation, JIN, for reliable fracture toughness evaluation with the 3PB specimens. The obtained values of JIN at room temperature and KIC at 77 K were 100– 120 kJ/m 2 and 20–22 MPam 1=2 , respectively, exhibiting little dependence on specimen size. By combining the experimentally obtained data with the plane-strain FEM analysis, a method was proposed to estimate JIN from a load-displacement curve measured for a single specimen. The method is applicable to heavily irradiated materials with little ductility.

Effect of heat treatment on mechanical properties and microstructures of Fe-8Cr-2W-0.1C-0.2V-0.04Ta martensitic steel F82H doped with about 60 mass ppm B or both of 60 mass ppm B and 200 mass ppm N has been examined. The normalization was heated at temperatures from 950 to 1250 °C for 1.8 ks, followed by air cooling or water quenching. After tempering treatment at 780 °C or 750 °C, the distributions of boron, boron nitride and oxygen were measured by a secondary ion mass spectrometry (SIMS). Optical microstructural observation and tensile and Charpy impact tests were performed also. In the boron doped F82H the tensile properties were similar to the non-doped F82H, but the ductile-brittle transition temperature (DBTT) shifted from -43 °C to 15 °C. SIMS images with high intensity of boron were observed in localized regions of the boron doped F82H. Water quenching reduced the DBTT shift, about 30°C, and the localized boron intensity was slightly decreased. In the boron and nitrogen doped tempered-F82H heat-treated by the water quenching from the normalizing temperature, the properties of tensile and Charpy impact were similar to the non doped F82H, and no pronounced localized boron image was observed in the SIMS image and no intensities of oxides and boron nitride were observed either.

The Japanese reduced activation ferritic steels (RAFSs) R&D road map towards DEMO is shown. The important steps include high-dose irradiation in fission reactors such as the high flux isotope reactor at Oak Ridge National Laboratory, irradiation tests with 14 MeV neutrons in the International Fusion Materials Irradiation Facility and application to ITER test blanket modules to provide an adequate database of RAFSs for the design of DEMO. The current status of RAFS development is also introduced. The major properties of concern are well-known, and process technologies are mostly ready for fusion application. RAFSs are now certainly ready to proceed to the next stage. A materials database is already in hand, and further progress is anticipated with the design of the ITER test blanket. Oxide dispersion strengthening steels are quite promising for high temperature operation of the blanket system, with potential improvements in radiation resistance and in corrosion resistance.

A Ti–35Al–10V intermetallic compound manufactured by powder metallurgy contains α2, β and γ phases. It has a better strength and ductility than the Ti–Al binary alloy containing α2 and γ phases. A typical 0.2% yield strength and total elongation of Ti–35Al–10V at 500 °C are 700 MPa and 15%, respectively. At 600 °C, the strength is still above 600 MPa and total elongation increases up to 60%. Transmission electron microscope (TEM) observation of the deformed microstructure suggests a transformation induced ductility in the β phase. After neutron irradiation of 3.5×1025 n/cm2 at 400 and 600 °C, the total elongation is only 10% at the 600 °C test, and almost no plastic elongation was observed at the 400 °C test. The TEM observation of irradiated Ti–35Al–10V did not show the formation of the ω phase.

The synergistic effect of displacement damage and hydrogen or helium atoms on microstructures in F82H steel irradiated at 250–400 °C to 2.8–51 dpa in HFIR has been examined using isotopes of 54 Fe or 10 B . Hydrogen atoms increased slightly the formation of dislocation loops and changed the Burgers vector for some parts of dislocation loops, and they also affected on the formation of cavity at 250 °C to 2.8 dpa. Helium atoms also influenced them at around 300 °C, and the effect of helium atoms was enhanced at 400 °C. Furthermore, the relations between microstructures and radiation-hardening or ductile to brittle transition temperature (DBTT) shift in F82H steel were discussed. The cause of the shift increase of DBTT is thought to be due to the hardening of dislocation loops and the formation of α′-precipitates on dislocation loops.

To make the best use of a limited number of irradiated strength-test specimens, it is desirable to derive the microstructure information related to the mechanical properties from mechanically tested specimens. A focused ion beam micro-sampling system was utilized to make thin film samples for transmission electron microscope (TEM) observations from the crack or fracture surface of the strength-tested specimens. The microstructure of mechanically tested specimens of a reduced-activation ferritic/martensitic steel, F82H, was investigated focusing on the irradiation effects on fatigue fracture. On unirradiated F82H, SEM observations revealed that the fatigue crack initiated along a prior austenite grain boundary. The TEM micrograph around the crack showed the presence of polygonization and the formation of distinctive shaped subgrains along the prior austenite grain boundary ahead of the crack. The fatigue crack initiated from an intergranular crack, and the TEM micrograph indicated that the main crack developed in the same manner as in the unirradiated F82H.

Solid hot-isostatic pressed (solid-HIPed) joint specimens and those with or without thermal-cycled specimens of 316LN-IG were irradiated at about 473 K to 1 dpa in the high flux isotope reactor. Slow strain rate tests were conducted in a high-purity, oxygenated ( dissolved oxygen =10 wt ppm) water at 423, 513 and 573 K with strain rates of (2–10)×10−7 s−1. Tensile tests were also conducted in vacuum at the same temperature range. Fracture surfaces were observed by scanning electron microscopy. No specimen showed irradiation-assisted stress corrosion cracking (IASCC) susceptibility at 423 and 513 K in water. At 573 K, however, intergranular cracks were observed to form in HIPed specimens. It was concluded that the effect of HIPing to IASCC susceptibility is small.

Effects of Al atoms on the hydrogen exchange reactions of hydrogen molecules with surface hydroxyls in silicates are investigated by ab initio calculations at the HF/6-31G ∗∗ and MP2/6-31G ∗∗ levels with the model clusters H3SiOH and H3Si(OH)Al(H)2OSiH3. The direct interaction of Al atoms with surface hydroxyl is found to bring about the lowering in the potential energy barrier of exchange reactions between H2 and H3SiOH. The lowering is explained by the changes of the reaction mechanism and the weakening of the O–H bond in surface hydroxyl by the interaction of Al atoms.

Type 316L(N) stainless steel of the international thermonuclear experimental reactor grade (316L(N)-IG SS) is being considered for the first wall/blanket module. Hot isostatic pressing (HIP) technique is expected for the fabrication of the module. To evaluate the integrity and susceptibility to stress corrosion cracking (SCC) of HIPed 316L(N)-IG SS, tensile tests in vacuum and slow strain rate tests in high temperature water were performed. Specimen with the HIPed joint had similar tensile properties to specimens of 316L(N)-IG SS, and did not show susceptibility to SCC in oxygenated water at 423 K. Thermally sensitized specimen was low susceptible to SCC even in the creviced condition. It is concluded that the tensile properties of HIPed SS are as high as those of the base alloy and the HIP process caused no deleterious effects.

The effect of simultaneous triple ion irradiation of He, H and Si on microstructural evolution of two kinds of SiC/SiC composites (HNS composite (using Hi-Nicalon type S SiC fiber) and TSA composite (using Tyranno SA SiC fiber)) at 1000 °C has been investigated. The microstructure observations of SiC/SiC composites irradiated to 10 dpa were examined by transmission electron microscopy. He bubbles were hardly formed in matrix of TSA composite, but many helium bubbles and some cracks were observed at grain boundaries of matrix of HNS composite. He bubbles and cracks were not, on the other hand, observed in the both fiber fabrics of HNS and TSA composites. Debonding between fiber and carbon layer following irradiation region was not observed in the both composites. Under these irradiation conditions, TSA composite showed the better microstructural stability against ion beams irradiation than one of HNS composite.

Tungsten-inert-gas weld joints prepared from the IEA heat of F82H were irradiated with 10.5 MeV Fe ions and 1.05 MeV He ions at 450 °C. Transmission electron microscopy observation revealed a marked cavity growth up to 30 nm at 50 dpa in the over-tempered portion of the heat-affected zone (HAZ), while cavities in the quenched portion of HAZ remained smaller (up to 10 nm). Base metal results also showed that a specimen tempered at 780 °C contained larger cavities than those tempered at 750 °C. Cavities in cold-worked specimens were the smallest. Initial dislocation densities in F82H, which are affected by heat treatment and/or mechanical treatment, dominate the cavity growth.

The optimization of the fabrication of SiC composites with stoichiometric SiC fibers (Hi-Nicalon Type S and Tyranno SA) was carried out by the forced thermal-gradient chemical vapor infiltration (FCVI) process. These SiC/SiC composites had a low porosity (11%) with uniform pore distribution and uniform thickness of carbon interphase between advanced SiC fibers and SiC matrix. The tensile strength was slightly increased with the thickness of the carbon interphase in the range of 75–300 nm. The effectiveness of the carbon interphase for the excellent mechanical properties was confirmed by scanning electron microscopy observation.

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.

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.

An overview of the present status of development of fusion nuclear technologies at Japan Atomic Energy Research Institute is presented. A tritium handling system for the ITER was designed, and the ...

We investigated tritium (T) release behavior from silica glass. The specimens were 8 kinds of commercially available silica glass. T was injected by the 6Li (n,α)T reaction of sintered pellets of lithium oxide (Li 2 O) into the silica glass with thermal neutrons in JRR-2 (VT-8) up to 5 × 10 18 neutrons/cm 2 at ambient temperature (about 350 K). After irradiation, the Li 2 O pellets were removed from the silica glass, and T release from the silica glass was measured in a flow of hydrogen (H 2 ) or ammonia (NH 3 ) sweep gas at atmospheric pressure at a constant heating rate of 2 K/min between 675 K and 1375 K with a proportional counter. In the case of H 2 sweep gas, a maximum tritium release rate was observed around 1023 K, while in the case of NH 3 sweep gas, two peaks around 1023 K and around 1123 K or a peak around 1123 K with a shoulder were obserbed. After the experiments of T release, FT-IR spectra showed a decrease of SiOH bands at 3650 cm -1 . On the other hand, no changes in intensities at 2250 cm -1 due to SiH were observed for both samples before and after T release.

The mechanism of the reaction of a water molecule with surface isolated hydroxyl of silica and silicates is investigated by ab initio molecular orbital calculations within a model reaction, H2O+H3SiOH, to figure out the role of water molecules in tritium release process. The hydrogen release from the surface isolated hydroxyl is found to proceed as a result of the Si–O bond breaking. Namely, tritium would be released by the hydroxyl-exchange reaction between water and surface hydroxyl of silicate, not by the hydrogen-exchange reaction. Prior to the exchange reaction, water molecules are found to prefer adsorbing as proton acceptors in the water – silanol complex, whereas the hydroxyl-exchange reaction occurs from a complex with a different form, in which the water molecule adsorbs as a proton donor. The overall potential energy barrier is calculated as 24.4 kcal mol−1 for the hydrogen release from the silicate in the presence of water molecules.

Effects of high-fluence irradiation in fusion-relevant helium production condition on defect cluster formation and swelling of reduced activation ferritic/martensitic steels (RAFs), JLF-1 (Fe–9Cr–2W–V–Ta) and F82H (Fe–8Cr–2W–V–Ta), have been investigated. Dual-ion (nickel plus helium ions) irradiation using electrostatic accelerators was adopted to simulate fusion neutron environment. The irradiation has been carried out up to a damage level of 100 displacement per atom (dpa) at around 723 K, at the HIT facility in the University of Tokyo. Thin foils for transmission electron microscopy (TEM) were prepared with a focused ion beam (FIB) microsampling system. The system enabled not only the broad cross-sectional TEM observation, but also the detailed study of irradiated microstructure, since unfavorable effects of ferromagnetism of a ferritic steel specimen were completely suppressed with this system by sampling a small volume in interests from the irradiated material.