A review is presented of recent progress attained in the understanding of the influence of transmutation on the development of fusion-relevant property change data. It is shown that early experiments on helium effects on void swelling, irradiation creep and tensile properties of austenitic stainless steels were often misleading, and the influence of helium is much smaller than originally expected. Similar definitive conclusions concerning the role of helium in ferritic steels, copper alloys, and vanadium alloys cannot be drawn at this time. It is also shown, however, that the formation of solid transmutants can be very important in a number of alloy systems, especially for some of those proposed as reduced radioactivation candidates.

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A composite model of the point defect processes and the extended defect evolution in irradiated materials was composed for a theoretical investigation of the response of fusion first wall materials to neutron bombardment. The point defects model calculates the concentration changes of single point defects including a transmutant helium, simple point defect clusters and complex clusters. The extended defect model consists of individual rate theory models of evolution of cavities, Frank faulted loops, network dislocation and other microstructural features. The model was calibrated based on the dual-ion experimental data on an Fe–15Cr–20Ni ternary austenitic model alloy. Using the calibrated model, the effects of helium generation on point defect processes, cavity nucleation, dislocation evolution and mechanical property change were investigated.

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The primary objective of the present study is to investigate the low temperature toughness of high Mn–Cr austenitic steels which are being developed as one candidate alloy for the structural material of fusion reactor from the point of view of reduced radio-activation. In the present work, the effects of V, Ti and P on the low temperature toughness of the alloys with the basic composition of 12%Cr-15%Mn-(0.1–0.2%)C-(0.1–0.2%)N are investigated. Coarse precipitates of M₂₃C₆ on the grain boundaries caused intergranular fracture and tiny precipitates of VN within the grains enhanced cleavage fracture in aging treated alloys. The addition of Ti was beneficial for improving toughness due to the restraining of the tiny VN precipitation. The addition of P, however, deteriorated the low temperature toughness of the alloys. The toughness values of the 12%Cr-15%Mn materials were compared with those of JFMS (a modified 9Cr–2Mo ferritic stainless steel) and JPCA (modified 316 austenitic stainless steel).

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The effects of phosphorus on microstructural evolution in austenitic stainless steels under fast neutron irradiation have been investigated. Three model alloys (Fe–16Cr–17Ni, Fe–16Cr–17Ni–0.024P and Fe–16Cr–17Ni–0.1P) were irradiated to 2 dpa and 11 dpa. The irradiation was performed in JOYO using SMIR-9. The nominal (calculated) irradiation temperatures were 673 (698), 773 and 873 K. Addition of phosphorus resulted in a suppression of void swelling at all temperatures except in 0.024P-alloy irradiated at 698 K to 11 dpa. The behavior of void swelling and phosphide formation by lower dose (2 dpa) irradiation was consistent with our previous studies of electron or high dose (40 dpa) neutron irradiations. In higher dose irradiation (11 dpa), on the other hand, void swelling and phosphide formation at 873 K were drastically suppressed. This may be caused by an unexpected temperature increase due to excess gamma-ray heating. A need for on-line temperature measurement and its active control are clearly demonstrated.

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R & D of ferritic/martensitic steels as structural materials for fusion reactor is one of the most important issues of fusion technology. The efforts to characterize microstructural evolution under irradiation in the conventional Fe–Cr–Mo steels as well as newly developed Fe–Cr–Mn or Fe–Cr–W low activation ferritic/martensitic steels have been continued. This paper provides some of the recent results of heavy irradiation effects on the microstructural evolution of ferritic/martensitic steels neutron irradiated in the FFTF/MOTA (Fast Flux Test Facility/Materials Open Test Assembly). Materials examined are Fe–10Cr–2Mo dual phase steel (JFMS: Japanese Ferritic/Martensitic Steel), Fe–12Cr–XMn–1Mo manganese stabilized martensitic steels and Fe–8Cr–2W tungsten stabilized low activation martensitic steel (F82H). JFMS showed excellent void swelling resistance similar to 12Cr martensitic steel such as HT-9, while the manganese stabilized steels and F82H showed less void swelling resistance with small amount of void swelling at 640–700 K (F82H: 0.14% at 678 K). As for irradiation response of precipitate behavior, significant formation of intermetallic χ phase was observed in the manganese stabilized steels along grain boundaries which is thought to cause mechanical property degradation. On the other hand, precipitates identified were the same type as those in unirradiated condition in F82H with no recognition of irradiation induced precipitates, which suggested satisfactory mechanical properties of F82H after the irradiation.

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A Fe–Cr–Ni model alloy was electron-irradiated using a high voltage electron microscopy (1000 kV), and in-situ observations on structural evolution and microchemical analyses were carried out. When the Fe–Cr–Ni alloy was irradiated, the nucleations of dislocation loops followed by voids were observed and at the same time when a grain boundary migration occurred. The compositional analysis after irradiation of an area including a grain boundary indicated nickel enrichment and chromium depletion near the grain boundary. It is suggested that when the radiation-induced point defects flow into the grain boundary, boundary migration and solutes redistribution are induced and the magnitudes depend on net point defects flow, especially that of interstitial atoms.

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A series of Fe–13Cr–15Mn alloys with different nickel levels of 0–15% have been examined by transmission electron microscopy and X-ray microanalysis following irradiation to 17–25 dpa at 693–823 K. All specimens were found to have developed typical features of radiation damage at high temperature; voids, dislocations and second phases. Both void formation and radiation-induced phase instability were found to be strongly dependent on nickel content. The ferrite phase was observed to form at grain boundaries in alloys with zero or low nickel content. Sigma phase often formed at the boundary between the ferrite and austenite phases. With increasing nickel content, both ferrite and sigma phase formation were suppressed. The induced density change was also found to be sensitive to the nickel content. The phase instability and density change during irradiation was explained in terms of radiation-induced solute segregation and formation of lower swelling phases.

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High Mn–Cr austenitic steels are still considered to be an important high temperature structural material from the point of view of reduced radio-activation. The objective of the present study is to make a fundamental research of mechanical properties and microstructure of 12%Cr-15%Mn austenitic steels. Especially the effects of alloying elements of V and Ti on the mechanical properties and microstructure evolution of high Mn–Cr steels were studied. Precipitation behaviors of carbides, nitrides and σ phase are investigated and their remarkable effects on the high temperature strength are found. The addition of V was very effective for strengthening the materials with the precipitation of fine VN. Ti was also found to be beneficial for the improvement of high temperature strength properties. The results of high temperature strengthes of the 12Cr–15Mn austenitic steels were compared with those of the other candidate and/or reference materials, for example, JFMS (modified 9Cr–2Mo ferritic stainless steel) and JPCAs (modified 316 austenitic stainless steels).

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In order to develop the small specimen technology in Charpy impact testing, the effects of V-notch dimensions on the test results were investigated for miniaturized specimens of a ferritic steel, Japanese Ferrite/Martensite Dual Phase Steel (JFMS). The miniaturized Charpy specimens had four different square cross-sections of 3.3, 2, 1.5 and 1 mm, and each of them had a variety of V-notch dimensions (notch depth, notch root radius and notch angle). All of the specimens were subjected to Charpy impact tests between 93 and 373 K using a specially instrumented impact machine. The fracture surfaces of all tested specimens were examined by scanning electron microscopy. The main results obtained are as follows:
(1) The ductile-to-brittle transition temperature (DBTT) varied noticeably depending upon the notch dimensions, some of the DBTTs exceeding that of the full size specimens. (2) The DBTTs for the miniaturized specimens were uniquely defined by the elastic stress concentration factor, K_t, leading to an important aspect that the DBTT for the full size specimens can be directly obtained from the DBTT of the miniaturized specimens with a V-notch giving a suitable value of K_t. (3) The upper shelf energy (USE) was dependent only on notch depth or ligament size, indicating that the notch geometry was practically unimportant. When all of the measured USEs were normalized by Bb² or (Bb)^3⁄2 (B is the specimen thickness, b the ligament size), the normalized USEs of the miniaturized specimens agreed with that of the full size specimens within the range of ±15% except for one specimen whose notch root radius was as large as 0.25 mm. (4) The observed characteristics of fracture surface were essentially the same as those of the full size specimens. The measurement of lateral expansion, or ductility, was more useful in estimating the impact property of JFMS than that of fracture appearance (fibrous fracture).

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The tensile properties of reduced-activation martensitic 9Cr–1WVTa and 9Cr–3WVTa steels for fusion reactors were investigated over the temperature range from room temperature to 873 K after neutron irradiation in the Japan Materials Testing Reactor at 538 K to fast neutron fluences of 2×10²³ and 3×10²⁴ n/m² (E>1 MeV). A conventional 9Cr–1MoVNb steel was also examined for comparison. The irradiation caused an increase in yield stress and a decrease in total elongation. Irradiation-produced defect clusters could not be seen directly by transmission electron microscopy. The increase in yield stress caused by irradiation Δσ_y was proportional to the 1/4 power of the neutron fluence. The Δσ_y was smaller in the 9Cr–1WVTa and 9Cr–3WVTa steels than in the 9Cr–1MoVNb steel and was smaller in the 1% W steel than in the 3% W steel. With increasing test temperature, the Δσ_y exhibited a further increase at 573∼673 K and then decreased to zero over the range from 673 to 873 K. The increase in irradiation hardening at 573∼673 K was postulated to be due to the formation of vacancy-carbon complexes, and the decrease in it at 673∼873 K was due to the annealing out of irradiation-produced defects. The decrease in total elongation was shown to be accompanied by a decrease in work hardening rate.

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Irradiation creep behavior of low activation steels, developed as structural materials for fusion reactors, was investigated. The objective of this study is to provide a fundamental understanding of the irradiation creep mechanism based on microstructural evolution under fast neutron irradiation. Pressurized tube creep specimens fabricated from tube segments were irradiated in the Fast Flux Test Facility (FFTF), Materials Open Test Assembly (MOTA) during FFTF Cycles 11 and 12. This paper provides the first creep results obtained after FFTF cycle-11 irradiation. (2.25–3)Cr–(1–2)W bainitic steels and 12Cr–2W ferritic/martensitic steels showed equivalent or superior creep resistance to a modified 316 stainless steel, known as Japanese Prime Candidate Alloy (JPCA), under fast neutron irradiation up to 600°C. For the case of ferritic steels, with increasing Cr content, creep strain increased up to 7 Cr and further increments of Cr content to 8, 9 and 12% reduced creep strain. Swelling enhanced creep near peak swelling temperature of 410°C was observed. Preliminary TEM observation suggests that irradiation induced precipitation and void nucleation were enhanced by applied stress near peak swelling temperature.

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Pressurized tube specimens of Japanese Ferritic-Martensitic Steel (JFMS) were irradiated in Fast Flux Test Facility (FFTF) Materials Open Test Assembly (MOTA) to 37.5 dpa at 680 and 793 K. Diametral creep strain following irradiation at 680 K was 0.11% even at zero hoop stress and slightly increased with increasing the hoop stress. The creep strain following irradiation at 793 K increased from zero to 0.23% with the increase in hoop stress from zero to 86 MPa. As for the void swelling, nothing was recognized. The martensitic phase in JFMS was stable after the irradiation at 680 K, but this was not the case at 793 K where a considerable recovery of dislocation structures was found. Following irradiation at 680 K, a high density of fine spherical G phase precipitates which were accompanied by the strain fields around them were observed, while large Mo-rich Laves phase particles were observed following irradiation at 793 K. Analysis of dislocation Burgers vectors revealed that dislocations having Burgers vector of type a⟨100⟩ were observed only after the irradiation at 680 K and that a large anisotropy was indicated in a⟨100⟩ type of Burgers vector populations for a stressed specimen, but no significant anisotropy was observed in the Burgers vectors of a specimen irradiated without hoop stress. The average diameter of small precipitates in the specimen irradiated at 680 K under stress was significantly larger than that without stress. Finally, the observed dimensional changes were attributed to both the precipitation-induced volume expansion and the Stress-Induced Preferential Absorption (SIPA) irradiation creep for 680 K irradiation and Climb-Controlled Glide (CCG) irradiation creep at 793 K.

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Recovery processes of martensitic transformation of neutron irradiated Ti-50.0, 50.5 and 51.0 at.%Ni alloys during post-irradiation annealing were investigated by means of differential scanning calorimetry (DSC), tensile tests and transmission electron microscope (TEM) observations. Neutron irradiation up to a fluence of 1.2×10²⁴ n/m² at 333 K suppressed the martensitic transformation as well as the stress-induced martensitic transformation of these alloys above 150 K. The TEM observations revealed that the disordered zones containing small defect clusters in high density were formed in the neutron irradiated Ti–Ni alloys. The DSC measurements also showed that the post-irradiation annealing caused recovery of the transformation of which the progress depended on the annealing temperature and period. A significant retardation of the recovery was recognized in the Ti-51.0 at.%Ni alloy in comparison with the Ti-50.0 at.%Ni alloy. From the shifts in the transformation temperature upon isothermal annealing at various annealing temperatures, the activation energies of the recovery process of the transformation in the neutron irradiated Ti-50.0 and 51.0 at.%Ni alloys were evaluated by a cross-cut method to be 1.2 eV and 1.5 eV, respectively. The recovery of the transformation was ascribed to the re-ordering resulting from decomposition of vacancy clusters, and those obtained values of the activation energy were considered to be the sum of the migration energy of vacancy and the binding energy of vacancy-vacancy cluster. The retardation of the recovery in the Ti-51.0 at.%Ni alloy was interpreted in terms of large binding energy in this alloy due to the off-stoichiometry.

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Effects of helium on the mechanical properties of vanadium base binary alloys have been studied using tritium trick technique. Emphasis was placed on the effect of atomic size factor of solute atoms. Substantial reduction of ductility was observed by tensile tests at room temperature by helium charging. Strong correlation was observed between apparent solid solution hardening, ductility reduction and atomic size factor of solutes. It is concluded that under-sized solutes strongly trap helium atoms. The distribution and the state of helium obtained by tritium trick technique has been suggested to be very much different from that anticipated for fusion neutron irradiation. Although tritium trick technique offers a unique and interesting way to study the effects of helium, the results should be taken with caution.

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Helium embrittlement is one of the problems in structural materials for fusion reactors. Recently, martensitic steels have been developed which have a good resistance to high-temperature helium embrittlement, but the mechanism has not yet been clarified. In this paper, tensile behaviors of helium implanted austenitic stainless steels, which are sensitive to the helium embrittlement, were studied and compared with those of martensitic steels under the same experimental conditions, and the effect of microstructure on helium embrittlement was discussed.
Helium was implanted by 300 appm at 573–623 K to miniature tensile specimens of 316 austenitic steels using a cyclotron accelerator. Solution annealed (316SA) and 20% cold worked (316CW) specimens were used. Post-implantation tensile tests were carried out at 573, 873 and 973 K. Yield stress at 573 K increased with the helium implantation in 316SA and 316CW, but the yield stress changes of 316SA at 873 and 973 K were different from that of 316CW. Black-dots were observed in the as-implanted specimen and bubbles were observed in the specimens tensile-tested at 873 and 973 K. Intergranular fracture was observed at only 973 K in both of the 316SA and 316CW specimens. Therefore, cold work did not suppress the high-temperature helium embrittlement under this experimental condition. The difference in the influence of helium on type 316 steel and 9Cr martensitic steels were discussed. Test temperature change of reduction in area showed clearly that helium embrittlement did not occur in 9Cr martensitic steels but occurred in 316 austenitic steels. Fine microstructures of 9Cr martensitic steels should suppress helium embrittlement at high temperatures.

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Hydrogen effects on the tensile properties of Type 316 stainless steel were investigated for the proposed International Thermonuclear Experimental Reactor (ITER) conditions of irradiation temperature (323∼523 K) and hydrogen concentration (∼2500 at ppm). While little changes in the tensile properties are observed in solution annealed Type 316SS for ITER conditions, a significant degradation of tensile ductility is recognized in the sensitized Type 316SS. The hydrogen induced ductility loss in the sensitized Type 316SS is accompanied by the brittle intergranular cracking and enhanced ductile rapture along grain boundaries at below and above room temperature, respectively. Microstructural observations by transmission electron microscope (TEM) revealed that the sensitization caused an increase in the density of Cr-carbides at the grain boundaries. High susceptibility to hydrogen embrittlement in the sensitized Type 316SS is attributed to the grain boundary precipitation relating phenomena such as the weakening of interface bonding between carbides and matrix and the martensitic transformation in the Cr-depleted zone near grain boundaries. Modified Type 316SS which contains 0.25 mass%Ti and named the Japanese prime candidate alloy (JPCA) shows good resistance to hydrogen induced intergranular cracking even after the sensitization treatment resulting from the suppression of Cr-carbides formation by means of the so-called stabilization of carbon by Ti. However, hydrogen still enhances the transgranular ductile rupture which is due to the formation of bubbles at Ti-carbides in the matrix of JPCA.
Based on the prediction of Cr-depletion and impurity enrichment associated with radiation-induced segregation (RIS) and radiation hardening following ITER relevant irradiation, hydrogen embrittlement of Type 316SS is assessed for the ITER conditions of irradiation temperature, hydrogen concentration and dose.

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The characteristics of the graphite as plasma facing material are reviewed and the problems of the graphite for the plasma performance are pointed out. The capabilities of low atomic number materials such as boron and beryllium are shown briefly and the limits in use of B or Be as the plasma facing component are presented. It is then summarized that the achievment of burning plasma condition with a long time period in a fusion experimental reactor such as ITER becomes difficult only by the development of the plasma facing materials.
In order to successfully achieve such burning plasma, a scheme with the enlarged divertor configuration is proposed for the reduction of the heat flux to the divertor. The present scheme is also compared with other schemes for the reduction of the heat flux.

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The defect structure and amorphization of 4.0 fJ (25 keV) D⁺ and He⁺ irradiated highly-oriented pyrolytic graphite have been systematically investigated in terms of irradiation dose and temperature. The graphite irradiated by either D⁺ or He⁺ is amorphized at a different critical dose, but the calculated critical dpa is the same in both cases, indicating no dominant chemical effect of the implanted particles. The damage structures appeared on the irradiated surface are characterized into the following three types; (a) lenticular openings, originating from gas accumulation in between the basal planes, (b) twins owing to the stress due to the depth dependent elongation, and (c) bubbles and blisters appeared after the amorphization. Raman spectroscopy distinguished three linear relationships between the peakwidth of 1580 cm⁻¹ and the peak intensity ratio of I₁₃₅₅/I₁₅₈₀. These relations respectively represent (1) accumulation of defects in the basal plane, (2) turbulence and disordering of the basal planes, and (3) amorphization accompanying relaxation of accumulated stress. The evolution of the defect structure and the mechanism of the amorphization are discussed in terms of the change of bonding nature and the stability of defect structures.

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A carbon fiber/SiC composite fabricated by a chemical vapor infiltration (CVI) process using ethyl-trichloro-silane (ETS) and methyl-trichloro-silane (MTS) as the sources of SiC at 1173∼1623 K was studied to develop a low activation material. Composites with a purity of better than 99.99% and a density of higher than 80% were obtained. The main matrix formed was β-SiC while silicon deposition also occurred for MTS. The mechanical properties were examined by the bending test at room temperature. Composites with a high strength of 800 MPa for ETS could be obtained. The fracture strength increased with decreasing thickness of the SiC layer, covering the surface of composites, and porosity. An apparent fracture toughness of the composite was 6∼10 MPa·m^1⁄2 which was about 3 times higher than that of monolithic SiC. The evaluation of induced activity of the composites was made assuming a first wall position of the fusion reactor. It was estimated that the γ-ray intensity decreases by about six orders of magnitude in a day and satisfies the allowable level of 25 μSv/h for personnel access by about 8-year cooling after the 10 MW·y/m² irradiation. This result shows that the present composites produced by CVI are potential low activation materials.

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High-purity molybdenum single crystals with the antitwinning axial orientation were irradiated with 14 MeV neutrons at room temperature, 473 K and 673 K under low neutron fluences of (2.1∼2.7)×10²¹ n/m² using the Rotating Target Neutron Sources (RTNS-II), and then were deformed in tension at 290 K. The irradiations led to the inhomogeneous yielding accompanied by large yield drops. The active slip plane, which of the unirradiated specimen was the plane (\bar101), was shifted more and more towards the maximum shear stress plane (\bar211) with decreasing irradiation temperature. By the isochronal annealing, the pronounced yield drop disappeared above 973 K. Furthermore, the flow stress of the specimens irradiated at room temperature changed from softening to hardening near 500 K and the strain-rate sensitivity also responded to this change, which was related to the migration of vacancies. Especially for the specimens irradiated at 673 K, it was found that the irradiation effects still remained above 973 K.

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Molybdenum alloys are candidate materials for high heat flux components of fusion reactors. In order to study the neutron irradiation damage at high fluence levels, disks of the molybdenum alloy TZM that had been stress relieved at 1199 K for 0.9 ks were irradiated in the FFTF/MOTA at 679, 793 and 873 K to a neutron fluence of 9.6×10²⁶ n/m² (E_n>0.1 MeV). Defect microstructures were observed by transmission electron microscopy. Dislocation structures consisted of isolated loops, aggregated loops (rafts) and elongated dislocations. The size of the loops increased with the irradiation tempetature. The levels of void swelling were 0.68 and 1.6% at 793 and 873 K, respectively. A void lattice was developed in the body-centered-cubic (b.c.c.) structure with a lattice spacing of 28 nm. The fine grain size (0.5–2 μm) was retained following high-temperature irradiation, indicating that the stress relief heat treatment may extend the material’s resistance to irradiation damage up to high fluence levels. The relationship between the microstructure and irradiation hardening was determined.

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In-reactor measurements of the spontaneous and long-term change in electrical conductivity of α-alumina were carried out in the JMTR fission reactor. The so-called radiation-induced conductivity, RIC, was observed and evaluated qualitatively as well as quantitatively. A long-term increase in electrical conductivity was observed, which is thought to be due to the so-called radiation-induced electrical degradation, RIED.

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Radiation damage in lithium orthosilicate (Li₄SiO₄) and Al-doped Li₄SiO₄ (Li_3.7Al_0.1SiO₄) irradiated with oxygen ions was studied with ionic conductivity measurements, Raman spectroscopy, Fourier transform infrared photo-acoustic spectroscopy (FT-IR PAS) and transmission electron microscopy. It was seen from the ionic conductivity measurements that lithium-ion vacancies were introduced as irradiation defects for Li-ions sites in both materials due to the irradiation. By the Raman spectroscopy, oxygen atoms in SiO₄ tetrahedra were considered to be preferentially displaced due to the irradiation for Li₄SiO₄, although only a decrease of the number of SiO₄ tetrahedra occurred for Li_3.7Al_0.1SiO₄ by displacement of both silicon and oxygen atoms. Decomposition of SiO₄ tetrahedra and formation of some new phases having Si–O–Si and Si–O bonds were found to take place for both Li₄SiO₄ and Li_3.7Al_0.1SiO₄ by FT-IR PAS. In the electron microscopy, damage microstructure consisting of many voids or cavities and amorphization were observed for Li₄SiO₄ irradiated with oxygen ions. The recovery behavior of radiation damage mentioned above was also investigated.

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The tritium surface reaction in lithium ceramics is investigated by combining in-situ tritium release experiments under neutron irradiation, out-of-pile experiments such as desorption studies, and modeling efforts. It was found that tritium is released from lithium ceramics in the chemical form of HT or HTO depending on the chemical composition of the sweep gas. The surface nature or existing state of surface hydroxyl group is important and was found to be affected by the sweep gas and sample pretreatment. The nature of the surface hydroxyl group also plays an important role in modeling the tritium release. In-situ observation of the surface hydroxyl group by infrared absorption spectroscopy revealed that the surface is heterogeneous for D₂O adsorption.

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