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In efforts to develop advanced tritium breeders, the effects of additives to lithium titanate (Li2TiO3) have been investigated, and good prospects have been obtained by using oxide additives such as TiO2, CaO and Li2O. As for the neutron multiplier, the development of a real-size electrode fabrication technique and the characterization of beryllium-based intermetallic compounds such as Be–Ti and Be–V have been performed. Properties of Be–Ti alloys have been found to be better than those of beryllium metal. In particular, steam interaction of a Be–Ti alloy was about 1/1000 as small as that of beryllium metal. These activities have led to bright prospects for the realization of the water-cooled DEMO breeder blanket by application of these advanced materials.

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 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.

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.

A binary pebble bed of lithium titanate (Li 2 TiO 3 ) was irradiated in the Japan Materials Testing Reactor (JMTR), and its tritium recovery characteristics bed was studied under pulsed neutron operations. The temperature at the outside edge of the pebble bed increased from 300 to 350 °C immediately after the window of hafnium (Hf) neutron absorber was turned toward the reactor core, while the tritium recovery rate increased gradually. The ratio of tritium recovery rate to generation rate at the high-power, ( R / G ) high , approached the saturated value of unity at about 20 h of operation. Overall tritium recovery behavior under the pulsed operation was similar to that under the steady state power operation. An estimated time constant of about 3 h for the tritium recovery was much longer than the thermal time constant of about 100 s.

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 disordering in the fused lithium metatitanate (Li 2 TiO 3 ) irradiated with high energy ions (160 MeV Xe, 80 MeV O) was examined by room-temperature X-ray diffraction and Raman spectroscopy. The exposure of Li 2 TiO 3 to Xe ions at fluences of 3.3 × 10 16 and 3.4 × 10 18 /m 2 was found to cause complete disordering in its surface layer, unlike the surface layer exposed to O ions up to maximum fluence of 1.4 × 10 20 /m 2 . An extent of disordering with irradiation is associated with the electronic stopping power of the incident ions in Li 2 TiO 3 rather than the fluence and accumulated radiation dose.

The engineering data on neutron irradiation performance of fusion blanket are indispensable to design the fusion blanket. In this study, an in-situ irradiation test of the Li 2 TiO 3 pebble bed was carried out at the J apan M aterials T esting R eactor (JMTR), and the effects of various parameters, i.e. sweep gas flow rate, irradiation temperature and hydrogen content in sweep gas on tritium release were evaluated.

Among the various tritium transport processes in lithium ceramics, the importance and the detailed mechanism of surface reactions remain to be elucidated. The dynamic adsorption and desorption model for tritium desorption from lithium ceramics, especially Li 2 O was constructed. From the experimental results, it was considered that both H 2 and H 2 O are dissociatively adsorbed on Li 2 O and generate OH − on the surface. In the first model developed in 1994, it was assumed that either the dissociative adsorption of H 2 or H 2 O on Li 2 O generates two OH − on the surface. However, recent calculation results show that the generation of one OH − and one H − is more stable than that of two OH − s by the dissociative adsorption of H 2 . Therefore, assumption of H 2 adsorption and desorption in the first model is improved and the tritium release behavior from Li 2 O surface is evaluated again by using the improved model. The tritium residence time on the Li 2 O surface is calculated using the improved model, and the results are compared with the experimental results. The calculation results using the improved model agree well with the experimental results than those using the first model.

The tritium release curves from the lower density region (71-86% TD) sintered pellets which were irradiated by thermal neutrons up to the fluences of 4 × 10 20 , 2 × 10 21 , 2 × 10 22 and 2 × 10 23 n/m 2 have one peak around 570, 590, 610 and 620 K, respectively. It means that no dependence on porosity is observed for these densities and fluences. This suggests that the tritium generated by thermal neutron irradiation is trapped in the irradiation defects introduced by irradiation of the bulk. The curves from the intermediate density region (87-89% TD) pellets which were irradiated up to same fluences as the lower also have one peak around 620, 640. 660 and 680 K, respectively. This means that the peaks tend to shift to higher temperatures by thermal neutron fluence, much the same as the 71-86% TD pellets. The curves from the higher density region (90-92.5% TD) pellets irradiated up to 4 x 10 20 n/m 2 have plural peaks. The curves from the pellets irradiated up to 2 x 10 21 , 2 x 10 22 and 2 x 10 23 n/m 2 also have the same number of peaks, and a dependence of the temperatures of the peaks on the fluences is not observed.

Depth profiles of implanted H atoms were measured for single crystalline Al2O3 samples irradiated at 923 K with dual or triple beams of 0.25 MeV H-, 0.6 MeV He-, 2.4 MeV O-ions or 2.6 MeV Al-ions. The peaks occur at 1.55 and 1.45 μm in the depth profiles measured for the H + Al dual beam irradiation and H + O dual beam case, respectively. The ratio of the peak areas is over 4, which is much larger than the implanted H atom ratio of 1.1, indicating that implanted Al atoms suppress the mobility of H atoms. However, the ratio becomes almost 1 between the triple beam samples with H + He + O-ions and with H + He + Al-ions at comparable doses. The fact demonstrates that implanted He atoms overwhelm the effects of the implanted self-cation/anion excess atoms on the migration behaviors of implanted hydrogen and radiation produced point defects, with the resulting sluggish cavity growth observed.

The world-wide efforts on ceramic breeder materials in the last two years concerned Li 2 O, Li 4 SiO 4 , Li 2 TiO 3 and Li 2 ZrO 3 , with a clear emphasis on the development of Li 2 TiO 3 . Pebble-manufacturing processes have been developed up to a 10 kg scale. Characterisation of materials has advanced. A jump-wise progress is observed in the characterisation of pebble-beds, in particular of their thermo-mechanical behaviour. Thermal property data are still limited. A number of breeder materials have been or are being irradiated in material test reactors like HFR and JMTR. The EXOTIC-8 series of in-pile experiments is a major source of tritium release data. This paper discusses the technical advancements and proposes a focus for further research and development (RD tritium release and retention properties; determination of the key factors limiting blanket life.

Pure copper was irradiated at 300–500°C by 5 MeV Cu ions (single beam) and Cu ions plus gas atoms (H and He) (dual beam irradiation) simultaneously. The high energy ion irradiation was carried out with the accelerator TIARA at the Takasaki-establishment of JAERI. The ions stop within a few microns from surface level and damage was formed up to this depth. The damage structure was observed as a function of the depth utilizing a focused ion beam (FIB) device. Below 300°C irradiation with a single beam produced a high density of stacking fault tetrahedra (SFT) but void formation was not observed. Large voids were observed with single beam irradiation at 500°C. In specimen irradiated with a dual beam of helium and Ni ions, the number density of voids was increased significantly. In copper irradiated with hydrogen and Ni ions, the number density of voids was not so large. Experimental results show that helium atoms promote void formation. Hydrogen atoms have less effect on void formation than helium atoms in pure copper.

A triple beam ion irradiation facility has been developed to study the synergistic effects of displacement damage, helium and hydrogen atoms on microstructural changes of materials under irradiation environments simulating a fusion reactor. The system consists of a vacuum chamber and three beamlines, which are connected with each electrostatic accelerator. Samples can be irradiated in the wide temperature range from liquid nitrogen to 1273 K in the chamber by replacing two kinds of sample stages alternatively. An austenitic stainless steel was simultaneously irradiated with triple beam of nickel, helium and hydrogen ions at 573–673 K using this facility and TEM observations were carried out from a cross sectional view normal to the incident surface. It was shown that the number density of dislocation loops decreased in the region where hydrogen and helium were deposited in comparison with ones in the region where only displacement damage was induced to a similar damage level.

Damage structures in a single crystal Al 2 O 3 sample irradiated at 923 K with simultaneous triple beams of H-, He-and O-ions with respective energies of 0.25, 0.9 and 4.7 MeV up to a fluence of 2.7, 3.5 and 5.5 x 10 20 m 2 , were examined by cross-section electron microscopy. High density dislocation loops are formed to a depth of 2.4 μm and the loop density increases with increasing depth. Cavities are observed in bands with the swelling peaked at depths of 1.4, 1.9 and 2.3 μm for H-, He- and O-ion implanted regions, respectively. The growth of cavities is most pronounced in the O-ion damaged region with a maximum size of 13 nm and a density of 2-3 x 10 23 m 3 . In a sample annealed for I h at 1273 K after the irradiation, cavities grown to an average size 40 and 60 nm with a maximum of 70 nm were observed in He- and O-ion implanted regions, respectively, and the density was decreased by about two orders.

Lithium oxide (Li2O) single crystals with various sizes (0.15–5 mm) were used as specimens. After the irradiation in JRR-4 and JRR-2 (thermal neutron fluence: 2 × 1017–2 × 1019 n/cm2), and fast neutrons in FFTF (fast neutron fluence: 4 × 1022 n/cm2), helium release from the Li2O specimens during the heating at a constant heating rate was continuously measured with a quadrupole mass spectrometer. The helium release curves from JRR-4 and JRR-2 specimens have only one broad peak each. From the dependence of the peak temperature on the neutron fluence and the crystal diameter, and the comparison with the results of sintered pellets, it is concluded that the helium generated in the crystal is released through the processes of bulk diffusion with trapping by irradiation defects such as some defect clusters. For the helium release from FFTF specimens, two broad peaks were observed in the release curves. It is considered that two different migration paths exist for helium migration in the specimen, that is, bulk diffusion and diffusion through the micro-cracks formed due to the heavy irradiation.

Li 2 TiO 3 single crystals with various grain size (1-2 mm) were used as specimens. After the irradiation up to 4 × 10 18 n/cm 2 with thermal neutrons in JRR-2, tritium release from the Li 2 TiO 3 specimens in isothermal heating tests was continuously measured with a proportional counter. The tritium release in the range from 625 to 1373 K seems to be controlled by bulk diffusion. The tritium effective diffusion coefficient (D T ) in Li 2 TiO 3 was evaluated to be D T [cm 2 /s]=0.100 exp(-104[kJ/mol]/RT), 625 K < T < 1373 K. In this temperature region, the tritium effective diffusion coefficients in Li 2 TiO 3 are close to those of Li 2 O irradiated with thermal neutrons of 4 × 10 16 and 2 × 10 19 n/cm 2 . It indicates that the tritium release performance of Li 2 TiO 3 is essentially as good as that of Li 2 O.

Lithium oxide (Li 2 O) pellets with various bulk densities (71, 80, 85 and 88% T.D.) were used as specimens. After the irradiation up to 2 x 10 17 -2 × 10 19 n cm -2 with thermal neutrons in JRR-4 and JRR-2, helium release from the Li 2 O specimens during a constant heating rate was continuously measured by a quadrupole mass spectrometer. In the helium release curves during a constant heating of 2 K min -1 , four kinds of peaks were observed. They are divided into two groups. One group was relatively wide peaks, of which the peak temperature range is 850-1100 K, appeared for all specimens. Another was a narrow and a very wide peaks, of which the peak temperature is around 1330 K, was remarkably observed only for the specimens with bulk density higher than 85% T.D. From the activation energy of the peaks and the neutron fluence dependencies of the temperatures of the peaks, the former group is considered to be attributed to the helium released through the processes of bulk diffusion, grain boundary diffusion without trapping at closed pores in the grains. On the other hand, from the bulk density dependence of the appearance of the peaks, the latter is considered to be attributed to the helium release with trapping at closed pores in the grains.

A cross-sectional electron microscopy technique was used to examine the microstructural evolution in single crystalline pure alumina (Al2O3) irradiated at 923 K with triple beams of 0.25 MeV H+, 2.4 MeV O2+ and 1.l MeV He+ to fluences of 6.3, 2.8 and 6.7 × 1020/m2. Irregularly shaped cavities are observed in two regions of depths of 0.2–1.8 and 2.2–2.5 μm. The cavity densities peaked at about 1.0 and 2.3 μm (7 × 1022/m3 and 4 × 1022/m3, respectively) with average sizes of 15 and 8 nm, respectively. The cavities formed in the He ion damaged region are aligned and mostly interconnected in the 〈0 0 0 1〉 direction to form irregularly shaped cavity tunnels. Tiny cavities, smaller than 5 nm, are formed inhomogeneously along irradiation-produced dislocations in the region 0.4–1.0 μm from the sample surface, where neither H nor O atoms are predicted to be stopped. This formation of cavities in the near-surface region is in marked contrast to that observed in the Al2O3 sample irradiated with simultaneous beams of 0.25 MeV H+ and 2.4 MeV O2+ ions to a fluence 2.5 times higher than that in the triple beam case. In the triple beam irradiation, cavities with comparable sizes are observed only near the end of ion ranges.

Damage structures in a single crystal α-Al2O3 sample irradiated with triple (0.25 MeV H+, 0.6 MeV He+ and 2.4 MeV O2+) ion beams at 923 K to a peak dose of 3.6 dpa were characterized by dislocation loops distributed from the region close to the incident surface to the region of 1.8 μm in depth, with cavities being formed in the restricted region of depths from 1.2 to 1.75 μm. A similar depth profile of damage structures was formed in a sample irradiated with dual (0.25 MeV H+ and 2.4 Me V O+) ion beams to a peak dose of 10.6 dpa, except that tiny cavities were formed down to the smaller depth of 0.5 μm. Depth profiles of implanted hydrogen atoms measured through a nuclear reaction analysis revaled that point defect mobilities were largely enhanced in the presence of hydrogen without helium. Upon annealing for 1 h at 1273 K, cavities grown to 70–80 nm at the maximum size were formed around peak damage regions in both the irradiated samples. Resultant cavity swellings were 6.2% and 18%, the ratio of which was found to be equal to that of accumulated damage.

An austenitic stainless steel was simultaneously irradiated with a triple beam of nickel, helium and hydrogen ions in the temperature range 573–673 K. Microstructural evolution in the specimen was observed from the cross-sectional direction normal to the incident surface using a transmission electron microscope. Synergistic effects of displacement damage, hydrogen and helium were characterized by decreasing the number density of dislocation loops in this temperature range.

The radiation-induced microstructural changes have been studied by cross-sectional transmission electron microscopy for single-crystal α-Al2O3 samples irradiated with triple ion beams (0.25 MeV H+, 0.6 MeV He+ and 2.4 MeV O2+; ‘Triple (A)’), (0.33 MeV H+, 0.45 MeV He+ and 1.3 MeV O+; ‘Triple (B)’) and three consecutive single ion beams (0.3 MeV H+ ion followed by 0.6 MeV He+ and then 0.8 MeV O+ ions) at 650°C to doses in the range 0.1–8.4 dpa at the damage peak. In the specimen irradiated with Triple (A), having the same average projected range to a total peak dose of 3.7 dpa, cavities with an average diameter of 13 nm were formed between 1.2 and 1.75 μm in depth causing a swelling of 0.1% at the peak, which is larger than those of the specimens irradiated with other conditions. The extent of the cavity-introduced region is some 40% smaller than observed in the damage region due to the He+ and the O+ ions and due to the H+ ions in the sample irradiated with Triple (B). Cavities formed in the sample irradiated with H+ ion beams were found to be inhomogeneously distributed in the vicinity of dislocation loops which were grown to loop sizes up to 80 nm.

Ceramic insulator materials used for various components such as diagnostic systems, in-vessel components, RF windows, etc. in fusion reactors are exposed to severe irradiation environments which are characterized by high energy neutrons with energies up to 14 MeV. The electrical resistivity of ceramic materials decreases due to radiation-induced conductivity (RIC) during irradiation. The RIC due to 14 MeV neutrons for Al 2 O 3 in the temperature range 300–570 K was measured in the neutron dose rate range 10 −2 to 10 0 Gy s −1 using the fusion neutronics source (FNS) at the Japan Atomic Energy Research Institute. The RIC of Al 2 O 3 was estimated for the more severe irradiation environment in ITER and a prototype fusion reactor (SSTR) by extrapolating the data due to 14 MeV neutrons in the present study and those due to gamma-ray in another study. The estimated electrical degradation due to the RIC is considered to be accommodated with appropriate fusion reactor designs.

Single crystal α-Al 2 O 3 specimens were irradiated with triple (0.25 MeV H + , 0.6 MeV He + and 2.4 MeV O 2+ ) ion beam, dual (0.25 MeV H + and 1.3 MeV O + ) ion beam and single ion beam (1.3 MeV O + ion irradiation followed by 0.25 MeV H + ion irradiation) at 650°C. In the specimen irradiated with the triple beam to a peak dose of 3.7 dpa, dislocation loops were distributed from the region very close to the surface to the depth of 1.8 μm, while cavities were formed at depths from 1.2 to 1.75 μm. Dislocation loops and cavities were not observed in the specimen irradiated with the dual beam. However, optical absorption spectroscopy for this specimen exhibited significant increase in intensity of absorption bands due to F and F + centers, which indicated production of oxygen vacancies by the irradiation. For the specimen subjected to the O + ion irradiation followed by the H + ion irradiation, no increase in the intensity of absorption bands was observed, although dislocation loops were densely formed at depths close to the projected ranges for each incident ion beams.

Modeling of the tritium reaction at the surface of Li2O was performed using data obtained in in-situ and out-of-pile tritium release experiments. In this model the effects of H2O and H2 in the swee...

The damage structures in Al/sub 2/O/sub 3/ after triple ion beam (H, He and/or N, O) irradiation were examined with cross-sectional TEM. Heavier ions penetration depths of He-and H-ions implantation enhanced growth of cavities produced by He- and H-ions, as compared with those produced by He-and H-ions only. Although peaks for He-implantation distribution are predicted by TRIM calculation to coincide with each other, cavity formation revealed the peaks to be separated appreciably.

The paper deals with the study of irradiation effects on the lithium metatitanate (Li/sub 2/TiO/sub 3/) ceramics. The simultaneous exposure of Li/sub 2/TiO/sub 3/ ceramics to H/sup +/, He/sup +/ and O/sup 2+/ ions was found to cause the appearance of anatase layer on its surface. The formation of the anatase is thought to be due to the displacement damage and electronic energy deposition. The irradiation of Li/sub 2/TiO/sub 3/ ceramics by 30 MeV O/sup 4+/ resulted in vanishing of grain structure near surface. This phenomenon is thought to be caused by the electronic energy deposition.

Pure copper and Fe-15Cr-2ONi alloy were irradiated at 500°C by 5 MeV Ni ions (single beam), Ni ions plus gas atoms (H and He) (dual beam irradiation) or triple beam of Ni, He and H ions (triple beam irradiation) simultaneously. Irradiation was carried out with accelerators TIARA at Takasaki-establishment of JAERI. Thin specimens were prepared with the combined technique of FIB (Focus Ion Beam) and electro-polishing. Damage structure versus the depth from surface was examined by TEM. In Fe-15Cr-20Ni alloy, voids were formed in the range of Ni ion irradiated specimens. In Fe-Cr-Ni alloy which is irradiated by a dual beam of Ni ion and He ion, the number density of voids increased significantly at the stopping range of He atoms. Remarkable results are void formation in specimens irradiated with dual beam of Ni and H ions, only voids of very small size were observed through the range of specimen. In irradiated specimen with triple beam, the size of voids was suppressed to be smaller than that of irradiated specimens with Ni and He ions. In pure copper which is irradiated with dual beam of helium and Ni ions, number density of voids were increased significantly. While in copper irradiated with hydrogen and Ni ions, number density of voids were not so large. Experimental results show that helium atoms have strong effect for void formation, while hydrogen atoms suppress void formation because the remarkable degradation of movement of small vacancy clusters can be induced by trapping hydrogen atoms.