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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 effects of cold working and impurities on swelling behavior in austenitic stainless steels irradiated at 400 °C to 17.3 dpa under spectrally tailored conditions in the Oak Ridge research reactor and high flux isotope reactor were investigated. The specimens were 20% cold-worked JPCA, 316R, K (low carbon (0.02%)) and C (low carbon (0.02%) and doped with 0.08% niobium). The helium generation rate was about 15 appm He/dpa. Cavities, dislocation loops and carbides were formed by irradiation in these steels. The swelling in the JPCA-CW and 316-CW was 0.003% and 0.004%, respectively and in the C-CW and K-CW was 0.02% and 0.01%, respectively. Swelling in K and C steels was strongly reduced by 20% cold-work, and the swelling in JPCA-CW and 316R-CW steels was comparable to JPCA-SA and 316R-SA steels. The synergistic treatments of addition of some impurities and cold working are very effective for the suppression of swelling at 400 °C in austenitic stainless steels.

Attractive characteristics of high-chromium iron-base alloys and chromium-base alloys have been demonstrated compared with those of conventional ferritic and austentic stainless steels. Recent progress especially on purification techniques indicates the possibility for these alloys to be developed into engineering materials. It is argued that the purification may overcome the brittleness of drawback of these alloys.

Tensile properties of high-purity Fe-50Cr and Fe-50Cr-8W alloys irradiated by neutrons have been investigated. The behavior of precipitate formation on dislocation loops in high-purity Fe-(9-70)Cr alloys has been examined during 1 MeV electron irradiation in a high voltage electron microscope. The neutron irradiation was performed at 673, 773 and 873 K in the Modified Japan Research Reactor-3 (JRR-3M) to about 0.4 dpa. In the Fe-50Cr and Fe-50Cr-8W alloys irradiated at 673 K, no elongation was observed in the test performed at 673 K. The elongations were observed to recover in the tensile test at 873 K. They also recovered largely after a heat treatment at 873 K for 10 h. However, heat treatment at 873 K for 10 h hardly improved the ductility for the Fe-50Cr alloy irradiated at 873 K. The defect clusters formed at 873 K by the irradiation were relatively stable. Moire fringe contrast due to the precipitate formation on dislocation loops was observed in the Fe-(9-70)Cr alloys irradiated by electrons at temperatures lower than about 773 K. The precipitates were identified as α'-phase in the Fe-(9-30)Cr alloys, but as α-phase in the Fe-70Cr alloy. The loops formed in Fe-(9-70)Cr alloys were of two types: either on {100} planes with Burgers vector b = a or on {111} planes with b = (a/2) . Their number density decreased with irradiation temperature.

Tensile and Charpy properties of low-activation ferritic steel, F82H irradiated up to 0.8 dpa at low temperature below 300°C were investigated. The helium effect on these properties was also investigated using the boron isotope doping method. Neutron irradiation increased yield stress accompanied with ductility loss, and it also shifted the ductile-to-brittle transition temperature (DBTT) from −50°C to 0°C. Boron-doped F82H showed larger degradation in DBTT and ductility than boron-free F82H, while they had the same yield stress before and after irradiation.

F82H(8Cr–2WVTa steel) IEA heat was used to prepare tungsten-inert-gas (TIG) and electron-beam (EB) weld joints, followed by heat treatment at 720°C for 1 h. Hardening in the weld metal and softening in the heat-affected zone (HAZ) were detected in TIG weld joints. In EB weld joints, hardening in the weld metal was more clearly observed but HAZ softening was hardly observed. Hardness of TIG weld metal was reduced after 550°C thermal-aging, but softening of the base metal was only observed after 650°C thermal-aging. M 23 C 6 phase was the major precipitate in aged base metal and weld joints. The amount of precipitates in aged weld metal was lower than that of normalized and tempered base metal. W-rich Laves phase was also detected in aged weld metal, HAZ and base metal.

Reduced-activation ferritic/martensitic steel, F82H (8Cr–2W–0.2V–0.04Ta–0.1C), and variants doped with isotopically tailored boron were irradiated at 673 K up to 51 dpa in the high flux isotope reactor (HFIR). The concentrations of 10B in these alloys were 4, 62, and 325 appm during HFIR irradiation which resulted in the production of 4, 62 and 325 appm He, respectively. After irradiation, transmission electron microscopy (TEM) was carried out. The number density of cavities increased and the average diameter of cavities decreased with increasing amounts of 10B. The number density decreased and the average diameter increased with increasing displacement damage. Swelling increased as a function of displacement damage and He concentration.

Four kinds of Ti–Al intermetallic compounds were made from powder metallurgical processing using mechanical alloying or plasma rotating electrode processing. One consisted of α2-Ti3Al single phase, and the others consisted of α2-Ti3Al and γ-TiAl duplex phases. These intermetallic compounds were irradiated at 673 K to the fluence of 5.16 × 1025 n/m2 (E > 1 MeV) in the high flux isotope reactor. After irradiation, transmission electron microscopy was carried out. Cavities were observed in both the α2-Ti3Al- and γ-TiAl-phases. The nucleation behavior of cavities in the α2-Ti3Al- and γ-TiAl-phases was influenced by chemical composition and fabrication processes.

A Ti–48at.%Al intermetallic compound has been irradiated with 200 keV He-ions at 623 and 773 K. The helium cavity density decreases with decreasing α2 and γ lamella width. A plot of the cavity density and lamella width reveals a linear relationship after irradiation to 15 dpa. Cavity density in regions with 300 nm wide lamella is about half of that in large γ grains of the specimen. The α2/γ lamellar boundaries supply a preferential nucleation site for cavities. Although the cavities are also formed on the γ/γ lamellar boundaries, the nucleation is limited to misfit dislocations on the boundaries. Defect-free zones are in the regions of about 50 nm width immediately adjacent to the lamellar boundaries at 773 K. These results suggest that the lamellar boundaries are effective sinks for radiation defects and contribute to the suppression of radiation-induced defect cluster development in TiAl alloys.

The effects of neutron spectrum on tensile properties of the low-activation martensitic steel F82H (8Cr–2WVTa) was examined using a thermal neutron shield to tailor the neutron spectrum for steels irradiated in the high flux isotope reactor (HFIR). The yield stresses of spectrally tailored specimens irradiated in HFIR to 5 dpa at 300°C and 500°C are on trend lines obtained from unshielded irradiation in HFIR. No significant effect of the neutron spectrum on tensile properties could be detected.

The synergistic effect of displacement damage and helium generation under neutron irradiation on tensile behavior and microstructures of austenitic stainless steels was investigated. The steels were irradiated at 400°C in the spectrally-tailored (ST) Oak Ridge research reactor/high flux isotope reactor (ORR/HFIR) capsule to 17 dpa with a helium production of about 200 appm and in the HFIR target capsule to 21 and 34 dpa with 1590 and 2500 appm He, respectively. The increase of yield strength in the target irradiation was larger than that in the ST irradiation because of the high-number density of Frank loops, bubbles, voids, and carbides. Based on the theory of dispersed barrier hardening, the strengths evaluated from these clusters coincide with the measured increase of yield strengths. This analysis suggests that the main factors of radiation hardening in the ST and the target irradiation at 400°C are Frank-type loops and cavities, respectively.

Helium was implanted into high purity Fe and F82H at room temperature up to 2000 appm to investigate helium effects on hardening. Ultra micro-indentation tests were performed on the specimens before and after helium implantation with loads that penetrate in 300 nm depth. After the indentation tests, the specimens were prepared with a focused ion beam (FIB) processing system for transmission electron microscopy (TEM) of the deformed regions. Results of the indentation tests indicated clearly that helium implantation caused hardening for both pure Fe and F82H. For pure Fe, it was also observed by TEM that the propagation of the plastic deformation zone formed during the indentation was limited to the helium-implanted layer, ranging from 600 to 800 nm from the incident surface.

Microstructural evolution in solution-annealed Japanese-PCA (JPCA-SA) and four other austenitic stainless steels, irradiated at 400°C to 17.3 dpa in the ORR and the high flux isotope reactor (HFIR) spectrally tailored experiment, were investigated by transmission electron microscopy (TEM). The mean He/dpa ratio throughout the irradiation fell between 12 and 16 appm He/dpa , which is close to the He/dpa values expected for fusion. In all the specimens, a bi-modal size distribution of cavities was observed and the number densities were about 1.0×10 22 m −3 . There was no significant difference between the number densities in the different alloys, although the root mean cubes of the cavity radius are quite different for each alloy. Precipitates of the MC type were also observed in the matrix and on grain boundaries in all alloys except a high-purity (HP) ternary alloy. The JPCA-SA (including 0.06% carbon and 0.027% phosphorus) and standard type 316 steel (including 0.06% carbon and 0.028% phosphorus) showed quite low-swelling values of about 0.016 and 0.015%, respectively, while a HP ternary austenitic alloy showed the highest swelling value of 2.9%. This suggests that the existence of impurities affects the cavity growth in austenitic stainless steels even at 400°C.

The tensile and impact behavior and microstructures of neutron-irradiated Fe-9Cr alloys with purity levels of 99.99 and 99.8 mass% have been examined. The neutron irradiation was mainly performed at about 255°C or 290°C to 0.3 dpa in the JRR-3M (Modified Japan Research Reactor-3). The increment of yield strengths in the high-purity (HP) and low-purity (LP) alloys induced by the irradiation was 225 and 170 MPa, respectively. The elongation in the HP alloy was remarkably reduced by the irradiation, and it was smaller than that in the LP alloy. The shift of the ductile-brittle transition temperature of HP alloy was about 175°C, and the shift was larger than that of the LP alloy. In TEM observations, dislocation loops were observed in HP and LP alloys. α'-like precipitates, about 6 nm in diameter, were observed on loops in the HP alloy, but in the LP alloy only dislocation loops were observed. Based on the theory of dispersed barrier hardening, the barrier strength factors of dislocation toops in HP and LP alloys are estimated to be about 0.4 and 0.3, respectively. The increase of the strength factor of loops in the HP alloy is thought to relate to the formation of the a'-like precipitates on the loops. From the relations between these mechanical properties and microstructures, the decrease in ductility of the Fe-9Cr alloy may be caused by the formation of α'-like precipitates on the dislocation loops.

Neutron-irradiation effects on the tensile properties of high-purity Fe-(9-50)Cr and Fe-50Cr-(0-8)W alloys were investigated. These alloys were irradiated at 400, 500, and 600°C to a damage of about 0.4 dpa in a reactor. The yield and ultimate strengths increased by the irradiation, and the increment of these strengths tended to increase with increasing chromium and tungsten contents. The tensile properties of the irradiated alloys were ductile, except for the Fe-50Cr and Fe-50Cr-W alloys irradiated at lower temperatures than 500°C. These behaviors are thought to be intimately related to α'-precipitates and dislocation loops formed by the irradiation.

The effect of radiation on microstructural evolution and phase transformation in Fe-50Cr-5W and Fe-50Cr-8W model alloys was examined. The 300keV Fe + ion irradiation was performed at temperatures of 550, 600, 670, and 700°C at a rate of 1.4 × 10 18 ions/m 2 s for (1.2-1.8) x 10 3 s in a transmission electron microscope (TEM). The damage peak level corresponded to about 190-285 dpa. The peak depth positions of damage and Fe + ion were about 60 and 100 nm, respectively. Dislocation loops were observed in the Fe-50Cr-W alloys up to 700°C. A Laves phase of tens of nanometers in size was formed in the Fe-50Cr-W alloys within a few minutes, but the formation of Laves phase decreased with increasing irradiation temperature. Sigma-phase grains of about 50-100 nm in size and micro-crystallines were observed in the alloys irradiated at 670 and 700°C. However, in the specimen aged at 700°C for 1000 h and the one aged at 700°C for 10 h after 80% CW, only the Laves phase had formed and no sigma phase was observed in the Fe-50Cr-W alloys. These results indicate that Fe + ion irradiation can strongly influence the Laves and sigma phase transformations, and microstructural development, in the Fe-50Cr-W alloys.

A Ti–40Al–10V (at%) intermetallic compound has been developed using vacuum arc remelting and hot-isostatic pressing (HIP), followed by isothermal hot-forging (IHF). The alloy, composed mainly of B2 and γ phases with equiaxial grains of several μ m in average diameter and a small amount of α 2 phase with equiaxial grains of smaller size, shows excellent tensile properties; it has an elongation larger than 6% on average and yield strength larger than 700 MPa at low (ambient temperature) to intermediate temperatures, although the strength decreases rapidly at temperatures higher than 600°C.

At temperatures below 703 K, irradiation increased flow stress and decreased elongation and fracture toughness in a 316 stainless steel. The residual ductility and fracture toughness after irradiation, however, were still high enough for structural applications. The effect of irradiation on the constitutive equation was evaluated. Results indicate that the alloy work hardens even after irradiation, and the residual work hardening capability is demonstrated to suppress flow localization. The relationship between residual fracture toughness and yield stress was examined, and irradiation effects on fatigue properties that cause channel fracture were analyzed.

In assessing possible potential issues for fusion applications, the compatibility of several metallic structural materials was examined using high temperature/pressure steam as test environment. High corrosion resistance associated with protective oxide film formation was regarded as essential for the function of protecting from tritium permeation and corrosion damage. A Ti–Al-based intermetallic compound with V addition, recently developed, showed excellent performance. A low-activation ferritic/martensitic steel, F82-H, was comparable with the current advanced materials for modern supercritical fossil boilers, while some potential vanadium alloys, although not intended for use in steam, were found less compatible.

Neutron irradiation-induced ductile-brittle transition behaviour of carburized molybdenum polycrystals and single crystals has been studied. These specimens were irradiated at 673, 873 and 1073 K to (7.9–9.8) × 10 23 n/m 2 ( E ⩾1 MeV). The results were analyzed as the difference in the behaviour between polycrystals and single crystals, and the effects of irradiation temperature. Three-point bend tests showed that the neutron irradiation-induced shift of ductile-brittle transition temperature (DBTT) at each irradiation temperature was much larger in single crystal specimens than in polycrystal ones. The carburization treatment was found to be more effective in polycrystals rather than in single crystals. This suggests that in polycrystals the grain boundary interface is strengthened due to carbon addition. At higher irradiation temperatures, this shift of DBTT decreased in both the specimens. It is characteristic in post-irradiation fracture behaviour than in single crystal specimens the cracks initiated from the island grains on the specimen surface.

A Ti-47 at.% Al intermetallic alloy and three TiAl alloys containing ∼2.0 at.% Mn and/or ∼0.4 at.% Si were prepared by powder metallurgical processing. When the samples were irradiated with He-ions to 3 dpa at 773 K, formation of defect clusters and cavities in TiAl alloys were remarkably suppressed by the addition of Mn. In Mn-added TiAl, although no loops, which were observed in pure TiAl and Si-added samples, were formed, the defect clusters with large strain field were found. It was suggested that the defect clusters were formed by the migration of mixed dumbbell type Mn atom-interstitials. The addition of Si showed no beneficial effects on suppression of radiation damage in TiAl alloys.

Microstructural evolution of welded austenitic stainless steel irradiated with mixed-spectrum neutrons was examined by transmission electron microscopy (TEM). TEM disks were obtained from electron-beam (EB) welded plates of JPCA, which is a Ti-midified austenitic stainless steel. Specimens were irradiated in HFIR up to 17 dpa at 670 and 770 K, and the estimated helium concentration was around 1100 appm. Cavities formed at 670 K irradiation were very small (

Diametral strain-controlled push–pull fatigue tests with zero mean strain were carried out with miniaturized hour-glass shaped specimens of an austenitic stainless steel in solution annealed condition at room temperature. The specimens had a diameter of 1.25 mm at the minimum cross section and a total length of 25.4 mm. The number of cycles to failure (Nf) was equal to or slightly greater than that obtained with standard size specimens. Nf was also revealed to be rather insensitive to the specimen load axis offset, indicating that the requirement of the specimen alignment to the load axis was not very severe for the miniaturized specimen.

The austenitic stainless steel (Fe–0.06C–0.5Si–1.8Mn–14Cr–16Ni–2Mo–0.24Ti) was irradiated in a triple ion facility and the High Flux Isotope Reactor. The materials used were in the solution annealed (SA) and 15% cold-worked (CW) condition. TEM and tensile specimens were irradiated to a dose level of 30 and 10 dpa at 200°C. Some of the specimens were annealed after the irradiation at 500°C for 8 h in a vacuum. Microhardness tests were carried out on the surface of the TEM disks at room temperature. Tensile tests were carried out at 200°C in a vacuum with strain rate of about 1×10−3 s−1. The microhardness of both SA and CW increased by ion irradiation and then decreased by annealing. The yield strengths of the neutron irradiated SA and CW decreased to 610 and 650 MPa by annealing, respectively. The strain to necking of the irradiated CW recovered from 0.7% to 7.6%. The fracture mode remained ductile in each case.

International research and development programs on reduced-activation ferritic/martensitic steels, the primary candidate-alloys for a DEMO fusion reactor and beyond, are briefly summarized, along with some information on conventional steels. An International Energy Agency (IEA) collaborative test program to determine the feasibility of reduced-activation ferritic/martensitic steels for fusion is in progress and will be completed within this century. Baseline properties including typical irradiation behavior for Fe–(7–9)%Cr reduced-activation ferritic steels are shown. Most of the data are for a heat of modified F82H steel, purchased for the IEA program. Experimental plans to explore possible problems and solutions for fusion devices using ferromagnetic materials are introduced. The preliminary results show that it should be possible to use a ferromagnetic vacuum vessel in tokamak devices.

TiAl-based intermetallic compounds with triple-phase equiaxial microstructure show excellent properties especially in ductility larger than 7% and strength larger than 700 MPa at low to intermediate temperatures. Studies on the irradiation behavior of these alloys have demonstrated good irradiation resistance, though much less information has been available concerning neutron-irradiation experiments. This recent progress gives them enormous potentialities not only for engineering materials, but also for nuclear application.

The tensile properties of high- and low-purity Fe–9, –18 and –30Cr alloys irradiated by neutrons up to a dose of 5 × 1024 n/m2 (E >1 MeV) at 613, 673, or 763 K have been examined. The yield strength and the ultimate strength are increased and the elongation is decreased by irradiation. The enhancement of these strengths due to the irradiation has a tendency to increase with chromium and impurity content. Large stress drops are often observed, especially at 763 K, in stress–strain curves of high-purity and high-chromium-content alloys except for Fe–9Cr alloys. Irradiation-induced precipitates, with 2% larger interplanar spacings than the α′-phase, on dislocation loops are more easily formed in the specimens of higher chromium content and higher purity. The precipitates are formed even in the irradiated Fe–9Cr alloy of high purity. The stress drop behaviour during the tensile tests is predominant in the specimens of higher chromium content and higher purity.

In order to clarify the effect of He atoms on the formation and annealing behavior of defect clusters in Ti–Al alloys, a Ti–47 at.% Al intermetallic compound has been irradiated with electrons and He-ions. Helium-ion irradiation enhances the nucleation of defect clusters, especially of interstitial loops, at temperatures from 623 to 773 K in both γ-TiAl and α2-Ti3Al grains of the sample. However, there is little difference between the annealing temperature ranges of defect clusters in TiAl grains formed by He-ion or electron irradiation at 623 K. The dot-shaped clusters and interstitial loops grow scarcely during annealing, but are annihilated by annealing up to 923 K. Cavities are formed after irradiation with He-ions below 10 dpa at 773 K, but no cavities are formed by electron irradiation up to 30 dpa. The cavities in γ-TiAl and α2-Ti3Al grains survive after annealing even at 1053 K for 1.8 ks, keeping their density and diameter to be nearly the same as those in the as-irradiated grains.

An accelerator-driven deuterium-lithium (d-Li) stripping reaction-type neutron source, such as the International Fusion Materials Irradiation Facility (IFMIF) planned by the International Energy Agency is recognized as one of the most promising facility to obtain test environments of high-energy neutrons for fusion reactor materials development. The limitation on the available irradiation volume of the irradiation facility requires the development of the small specimen test techniques (SSTT). Application of SSTT to evaluate the degradation of various components in the light water reactor for the life extension is expected to be also quite beneficial.A remote-controlled testing machine for the Small Punch (SP) and miniaturized tensile tests was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR). The machine is designed for testing at temperatures ranging between 93 and 1, 123K to evaluate the temperature dependence of the strength of materials including the embrittlement at low temperatures and the softening at elevated temperatures. The tests are performed in a vacuum or in an inert gas environment.The machine has been installed in a hot cell and is being used for the round robin test program of the SP test method. The round robin test program is planned to identify the capability of the test method and to establish a standard test procedure. The configuration and the specifications of the test machine are introduced and the results of the SP tests are also shown.

TiAl intermetallic alloys with ordered structures for nuclear applications are examined based on recent experimental results. TiAl intermetallic alloys are better irradiation resistance compared with that of metallic materials like austenitic stainless steels; no degradation in tensile properties and very slow development in microstructure under irradiation including swelling. The materials with these attractive properties along with inherent physical ones are one of the candidate materials for the next generation nuclear machines, although these are not yet matured as engineering materials.

Vanadium, V1Ni, V10Ti and V10Ti1Ni (at%) were irradiated in HFIR to doses ranging from 18 to 30 dpa and temperatures between 300 and 600°C. Since the irradiation was conducted in a highly thermalized neutron spectrum without shielding against thermal neutrons, significant levels of chromium (15–22%) were formed by transmutation. The addition of such large chromium levels caused strong embrittlement. At higher irradiation temperatures radiation-induced segregation of transmutant Cr and solute Ti at specimen surfaces caused strong increases in the density of the alloy. The resultant shrinkage, possibly compounded by thermal cycling, led to cracks developing at all intersections of grain boundaries with the specimen surface. This caused specimens irradiated at 500°C or below to often fail during retrieval from the reactor, as well as during electropolishing and other handling operations. At 600°C, the cracking and embrittlement processes are so severe that only a fine dust, composed mostly of individual grains or chunks of grains, was found in the irradiation capsule.

This paper reviews the history and the present status of the development of low-activation ferritic/martensitic steels for fusion applications, followed by a summary of the status of the International Energy Agency fusion materials working group activities, where an international collaborative test program on low-activation ferritic/martensitic steels for fusion is in progress. The objective of the test program is to verify the feasibility of using ferritic/martensitic steels for fusion by an extensive test program covering the most relevant technical issues for the qualification of a material for nuclear application. The development of a comprehensive data base on the representative industrially processed reduced-activation steels of type 89Cr2WVTa will provide designers a preliminary set of material data within about 3 years for the mechanical design of components, e.g., for demo relevant blanket modules to be tested in ITER. Knowledge on the current limitations of low-activation ferritic steels for application in advanced fusion systems is reviewed and future prospects are defined.

Tensile and Charpy impact properties of neutron irradiated F82H (Fe8Cr2WVTa) with and without boron have been investigated to obtain the basic irradiation response on mechanical properties in low damage regime less than 1 dpa at the temperature ranging from 300° to 590°C. Boron-doped steel was used for the helium effect due to (n, α) reaction. Typical irradiation hardening was observed at 300°C. The irradiation above 520°C did not reveal increase in yield stress, but the specimen irradiated at 590°C showed some reduction in elongation in room temperature tensile testing. Slight difference in the tensile properties between boron-doped and boron-free were observed at 590°C. No changes in ductile brittle transition temperature (DBTT) occurred at a temperature between 335° and 460°C by Charpy impact testing.

J—R curves of an austenitic stainless steel in solution annealed and cold worked conditions were obtained using miniaturized fracture toughness specimens and standard compact tension specimens. Results indicate that the specimen size effect for the cold worked steel was small. JQ values of irradiated miniaturized specimens agreed well with those of cold worked specimens with similar yield stress levels. This suggests that the irradiation induced degradation of the fracture toughness is mainly dependent upon the irradiation hardening.

In a candidate austenitic stainless steel (316F) for fusion reactor structural materials, irradiation behavior of the weld metal produced by electron-beam welding (containing 7.9 vol% δ-ferrite) was investigated in terms of microstructural development. The densities of interstitial clusters in the γ-phase of the weld metal irradiated with He-ions at 673 and 773 K were about four times larger than those in 316F. Voids were formed in the δ-ferrite of the weld irradiated at 773 K. The number of clusters decreased in the weld metal (γ-phase) aged at 773 to 973 K, compared with that in the as-welded metal. The change in cluster density could be attributed to a Ni concentration increase in the γ-phase of the weld metal during aging.

The post-irradiation weldability of neutron-irradiated type 316 stainless steel has been examined. The tungsten inert-gas arc welded joints were made of type 316 stainless steel used as wrapper tube in the JOYO fast reactor at 668–683 K to a maximum dose and helium content of about 22 dpa and 9 appm respectively. Large loss of tensile ductility was observed in welded joints made of the irradiated steels, being accompanied with intergranular brittle fracture in the heat affected zone at ambient temperature (about 293 K) and at 773 K, although ductile fracture of unirradiated specimens was observed in the weld metal region. The brittle fracture is closely related to helium bubble formation on grain boundaries during the welding.

Post-irradiation tensile properties of Ti-rich TiAl (approximately Ti-47at% Al) intermetallic alloys, produced from the powder product manufactured by a plasma rotating electrode process (PREP), are investigated. The TiAl alloy powders are hot isostatic pressed (HlPed) into a cylindrical large compact, and then consolidated and formed by isothermal hot-forging (IHF). The final duplex structure of the alloys consists of equiaxial grains of the Ti3Al and TiAl(γ) phases with a grain size of approximately 2 μm in diameter. An increase in ductility from 6.0 to 10.3% with a slight increase in yield strength, is observed in specimens tested at 873 K after neutron irradiation at 873 K to a dose of 1 × 1020 n cm−2 (E>1 MeV) in JRR-2. This phenomenon can be understood by a mechanism that involves twins, which are believed to play an important role in the deformation process, nucleating during irradiation and growing during tensile testing after irradiation.

Addition of 1 at% nickel to vanadium and V-10Ti, followed by irradiation along with the nickel-free metals in HFIR to 2.3x1026 n m−2, E > 0.1 MeV (corresponding to 17.7 doa) at 400°C, has been used to study the influence of helium on microstructural evolution and embrittlement. Approximately 15.3% of the vanadium transmuted to chromium in these alloys. The ∼50 appm helium generated from the 58Ni(n, γ)59Ni(n, α56 Fesequence was found to exert much less influence than either the nickel directly or the chromium formed by transmutation. The V-10Ti and V-10Ti-1Ni alloys developed an extreme fragility and broke into smaller pieces in response to minor physical insults during density measurements. A similar behavior was not observed in pure V or V-1Ni. Helium's role in determination of mechanical properties and embrittlement of vanadium alloys in HFIR is overshadowed by the influence of alloying elements such as titanium and chromium. Both elements have been shown to increase the DBTT rather rapidly in the region of 10% (Cr + Ti). Since Ci is produced by transmutation of V, this is a possible mechanism for the embrittlement. Large effects on the DBTT may have also resulted from uncontrolled accumulation of interstitial elements such as C, N, and O during irradiation.

Microstructural evolution under heavy-ion irradiation of the ferrite/austenite duplex stainless steel (composition: 0.019C-0.49Si-1.01Mn-9.0Ni-21.3Cr-2.5Mo-65.7Fe in wt%) was studied. Irradiation was performed using 4 MeV nickel ions at the high-fluence irradiation test (HIT) facility at 773 K to a dose of 10 dpa at peak. A conventional transmission electron microscope (TEM) and a TEM with a field emission gun were used to observe the microstructure. A high density of dislocation lines and loops and a few small voids were observed in the austenitic phase. On the other hand, a moderate density of dislocation lines and the bcc χ-phase were found in the ferritic phase. There were no voids in it. The χ-phase has a lattice parameter of 0.86 nm and its chemical composition was 4Si-13Mo-27Cr-36Fe-19Ni (in wt%). This phase appears to be radiation-induced during heavy-ion irradiation.

Electron beam (EB) welding was applied to 316SS and the titanium modified 316SS (JPCA). For the prospective improvement of swelling in welded joints, modified TIG or EB welding procedures utilizing titanium or nickel foil insertion were employed. For the case of EB welding of 15 mm thickness I-butt joint, the higher weld heat input showed better swelling resistance in the joints. The in-reactor creep results suggest that irradiation creep in welded joints may not be a big concern, as far as swelling resistance is maintained. So, Ni addition, stress relief treatment and high heat input for EB welding with optimization of welding condition are recommended for suppressing irradiation creep and swelling.

The optimization of a stable austenitic F5 (18Cr-35Ni-Mo) steel was carried out by means of electron beam refining and the SAR treatment with a scavenger element Ti. Excellent irradiation resistance was confirmed by simulation tests of the irradiation induced degradation that is controlled by interactions between supersaturated solutes and crystal defects like dislocations.

Pure and Cr2O3-doped single crystal α-Al2O3 specimens were irradiated with 330 keV O+-ions, 400 keV He+-ions and 120 keV electrons at temperatures of 1123 to 1323 K, and the damage structure was studied by transmission electron microscopy. Dislocations, cavities and γ-Al2O3 grains recrystallized from α-Al2O3 were formed by O-ion irradiation at 1223 K. The damage region extended to a depth of about twice as large as the peak damage depth predicted by TRIM85, and cracking occurred in the deeper part of the region. Recrystallization and cracking were not observed in the specimen irradiated with He-ions up to nearly the same dose as the case of O-ion irradiation. Instead, dislocation loops and cavities were observed. Damage near surface on basal planes occurred during 120 keV electron irradiation at 1273 K in the He preirradiated pure Al2O3 specimen.

Microstructural evolution during helium-ion irradiation at 773 K in a weld metal sample (containing 10% δ-ferrite) of Ti-modified austenitic stainless steel was observed in-situ using a transmission electron microscope. Very fine helium bubbles formed at high number density in both δ-ferrite and austenite by a dose of 3 × 1019ionsm−2. However, the different microstructural evolution developed in the two phases with increasing dose. Fine bubbles in δ-ferrite rapidly grew with increasing dose and coalescenced at doses beyond 9 × 1019ionsm−2. Tiny bubbles in austenite remained very stable during irradiation to a dose of 2 × 1020ionsm−2. The number density of bubbles was about one order of magnitude larger in austenite than that in δ-ferrite, and increased with increasing dose. Swelling became much larger in δ-ferrite than in austenite, as a result. This is the inverse phenomenon to the conventional result that swelling is lower in ferrite than in austenite. Sigma phase formed by radiation-enhancement at grain boundaries between δ-ferrite and austenite and at the interfaces within δ-ferrite at a dose 9 × 1019ionsm−2 and grew with increasing dose. The chemical composition of σ-phase formed during irradiation showed Cr and Mo enrichment, and Fe and Ni depletion compared with σ-phase formed thermally.

Damage structure in a Ti-rich TiAl intermetallic compound has been investigated by transmission electron microscopy after He-ion irradiations up to 3 × 10 21 ions/m2 at 623 and 773 K. The irradiations resulted in loop-shaped or dot clusters in the γ-TiAl and α 2 - Ti 3 Al grains. The cluster density in TiAl was over one order of magnitude lower than that in Ti3Al. The clusters in TiAl irradiated to 3 × 10 21 ions/m2 at 773 K were identified as interstitial-type faulted loops lying on {111} planes with the Burgers vector direction of 〈111〉. Cavities were created along the grain boundaries as well as in the matrix in TiAl at 773 K. The EELS analysis indicated that the cavity formation was associated with injected He atoms.

The structure of a mechanically alloyed Ti-45at%Al powder and its decomposition behavior during sintering have been investigated by X-ray diffraction (XRD) and analytical transmission electron microscopy (TEM). By the XRD experiments, the powder after 200 hr mechanical alloying is shown to be consisted of two phases, Ti3A1 and TiAI. The volume fraction of Ti3Al is estimated to be larger than that of TiAI, while for the sintered compact of the powder this tendency was reversed. The mechanically alloyed powder is composed of strain free fine grains (grain size ranging from 10 to 20 nm in diameter). The average composition of the grains are approximately Ti-45at%Al. Therefore, the powder may be assumed to be composed mainly of Al supersaturated Ti3Al. The results of XRD and analytical TEM experiments suggest that the decomposition of the supersaturated Ti3AI to equilibrium Ti3Al and TiAI took place during sintering.