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The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokka...

Understanding of Li mass transfer behavior in a tritium breeding blanket is an important issue from viewpoints of establishment of tritium cycle and tritium safety. In this work, weight reduction property of Li2TiO3 with excess Li pebbles, which were fabricated by National Institutes for Quantum and Radiological Science and Technology, was investigated and the amount of Li mass loss and the rate of Li mass loss in water vapor atmosphere at elevated temperatures were evaluated. The Li mass loss proceeding at 900 °C with a relatively high rate was limited. The Li mass loss of Pebble210 (Li/Ti = 2.10) was 1.2 wt% and that of Pebbe211 (Li/Ti = 2.11) was 1.4 wt%, eventually. The rate of the Li mass loss increased with increasing temperature and it seemed to increase proportionally to a square root of water vapor pressure. An empirical formula for the Li mass loss was proposed as a function of temperature, heating time and water vapor pressure in a purge gas.

New tritium breeding materials that maintain large tritium breeding ratio for thermonuclear fusion reactors should be developed. In this study, we investigate the lithium-vaporization behavior from the sintered bodies of Li8ZrO6 and Li5AlO4 having high lithium atom densities. From the viewpoint of lithium density, we investigate the possibility of using the sintered compacts of Li8ZrO6 and Li5AlO4 as a tritium breeder. Lithium vaporization rates from the sintered bodies of Li8ZrO6 and Li5AlO4 are high. However, for both materials, the lithium vaporization rate gradually decreases with the annealing time by “self-stabilization” in which the surfaces of each sintered body are covered by the stable layer of the second phase. Considering the temperature dependency of equivalent vapor pressure and the volume fraction of tritium breeding material loaded at high-temperature region, it is suggested that the sintered compacts of Li8ZrO6 and Li5AlO4 have high potentials for the tritium breeder.

The lithium isotope effect in ionic superconductors is numerically investigated. Using first-principles and kinetic Monte–Carlo simulations, the diffusion coefficients of 6Li and 7Li ions are calculated. Two important effects that cause the difference in the mobilities of lithium isotopes are also shown: the quantum effect, which decreases the activation barrier energy for lithium ion in a certain structure and increases the ratio of the diffusion coefficient of Li isotopes, as well as the correlation effect between 6Li and 7Li ions and vacancies. These results will assist in finding materials that exhibit large lithium isotope effect. The preferred chemical composition of lithium isotopes required to enhance the lithium isotope effect and the application of these isotopes as a separation membrane for the lithium ionic superconductors are discussed.

Solid tritium breeding materials are expected to be used in high-temperature conditions for a long time in a fusion DEMO reactor. Thus, it is important to understand how bred tritium releases from the long-term heated material under the environment as close to a DEMO condition as possible to establish a tritium fuel cycle and keep it safe. In this work, the tritium release behavior from the irradiated Li2TiO3 pebbles that was preheated for 720 h at most was observed by heating at 1000 °C under 1000 Pa H2/Ar gas flow. The release peaks of HTO and HT were observed around 300 °C due to the desorption of the chemisorbed water. Also, the broad HTO peak was observed in a higher temperature region despite purging H2/Ar gas. This result suggests that this tritium was released without exchanging with H2 but with combining with oxygen in the pebbles. Moreover, the released tritium amount decreased as the pre-heating time. Finally, the amount of tritium that could not be released by the heating experiment was quantified by dissolving samples with an acid solution. Besides, the total tritium release ratio was discussed.

Understanding of Li evaporation properties is important because Li mass transfer decreases tritium breeding ratio and influences tritium behavior possibly. In QST, the pebbles of Li2TiO3 with excess Li were fabricated by the optimized sintering condition recently and it was confirmed that the pebbles have a good tritium release property. In this work, the weight reduction of the pebbles at elevated temperatures was investigated considering the releases of water vapor and CO2 from the pebbles. Furthermore the mechanism of Li mass loss from Li2TiO3 with excess Li pebbles was discussed by comparing the difference between the pebbles formed by the optimal fabrication method, Pebble210 and ones formed by the previous fabrication method, Pebble211. The release behaviors of water vapor from Pebble210 and Pebble211 were almost same except a small peak at 350 °C. A large amount of CO2 was released from Pebble211 at over 700 °C which is close to the melting point of Li2CO3. On the other hand, the release of CO2 from Pebble210 was not observed. The weight reduction excluding the contribution of water vapor for Pebble210 was smaller than that for Pebble211. The difference of weight reduction between Pebble210 and Pebble211 are caused by the presence of Li2CO3.

Changes in mechanical property and microstructure of sintered pebbles of lithium metatitanate (Li2TiO3) and Li excessive lithium metatitanate (Li2+xTiO3, x > 0) as tritium breeder pebbles for thermal fusion reactors, which were caused by thermal annealing, were investigated. The compressive strength was measured by using a universal tester. The Li2TiO3 pebble was stable during the thermal annealing at 900 °C, and its compressive strength was not decreased. In contrast, the microstructure and Li/Ti ratio of the Li2+xTiO3 pebble were markedly changed during the thermal annealing; the grain growth was severe and the excessive Li was disappeared in a 1000 h annealing. As a result, its compressive strength was significantly changed; it was drastically decreased by a 100 h annealing but it was once increased again in the additional annealing by a change in fracture mode.

Lithium vaporization properties of sintered compacts of single-phase Li2TiO3, Li8ZrO6, β-Li5AlO4, and two-phase β-Li5AlO4-Li8ZrO6 and β-Li5AlO4-Li2TiO3 are investigated. Li vaporization rates from the sintered pellets are measured by gravimetric methods, and crystalline phase changes are analyzed by X-ray diffraction. Li2TiO3 is stable during thermal annealing at 900 °C. In contrast, Li is evaporated from the other materials, and β-Li5AlO4 and Li8ZrO6 are, respectively, decomposed to LiAlO2 and a mixture of Li6Zr2O7 and Li2ZrO3. Considering the high Li density, which is estimated from the Li vaporization rate and initial Li density in the pebble, and appearance stability, the sintered compacts of single-phase β-Li5AlO4 and the two-phase materials β-Li5AlO4-Li2TiO3 and β-Li5AlO4-Li8ZrO6 have the potential to be applied to new advanced tritium breeders with high Li density.

Several key components, such as superconducting coils, remote handling equipment, heating systems, have been designed and manufactured by JADA (Japan Domestic Agency). These activities have been carried out in accordance with the agreed schedule; in collaboration with the ITER organization and other domestic agencies. As a significant technical program using ITER, to design and to manufacture the TBS (Test blanket system), some R&D and design activities have also been conducted in Japan. Under the IFERC (the International Fusion Energy Research Center) projects of BA (Broader Approach) activities, design and R&D activities on fusion DEMO reactor have been carried out. For the DEMO R&D activity, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. From 2007, the above project...

The pebble properties of novel two-phase Li 4 SiO 4 pebbles of 1 mm diameter with additions of Li 2 TiO 3 , LiAlO 2 or Li x La y TiO 3 are evaluated in this work as a function of the second phase concentration and the microstructure of the pebbles. The characterization focused on the mechanical strength, microstructure and open as well as closed porosity. Therefore crush load tests, SEM analyses as well as helium pycnometry and optical image analysis were performed, respectively. This work shows that generally additions of a second phase to Li 4 SiO 4 considerably improve the mechanical strength. It also shows that the fabrication processes have to be well-controlled to achieve high mechanical strengths. When Li 2 TiO 3 is added in different concentrations, the determinant for the crush load seems to be the open porosity of the pebbles. The strengthening effect of LiAlO 2 compared to Li 2 TiO 3 is similar, while additions of Li x La y TiO 3 increase the mechanical strength much more. Yet, Li 4 SiO 4 and Li x La y TiO 3 react with each other to a number of different phases upon sintering. In general the pebble properties of all samples are favorable for use within a fusion breeder blanket.

Wet-chemical fabrication processes are highly adaptable to a wide range of raw materials and are therefore well suited for evaluating new material compositions. Here the established emulsion method was modified to fabricate novel two-phase Li 4 SiO 4 pebbles of 1 mm diameter with additions of Li 2 TiO 3 , LiAlO 2 or Li x La y TiO 3 . As the lithium density of the latter two compounds is relatively low, only moderate contents were added. The Li 2 TiO 3 additions, however, cover the full compositional range. The fabrication process was characterized with regard to its constancy and aptness for the anticipated pebble compositions by optical pebble size measurements. Also the phase content and the elemental composition of the fabricated pebbles were analyzed by XRD and ICP-OES combined with XRF, respectively. This work shows that the emulsion method is an appropriate method to produce pebbles with the anticipated Li 2 TiO 3 and LiAlO 2 concentrations in a Li 4 SiO 4 matrix. However, Li 4 SiO 4 and Li x La y TiO 3 react with each other to a number of different phases. To evaluate the activation properties of the pebbles, FISPACT calculations with a DEMO relevant neutron source are applied as well. The addition of aluminum seems to be unfavorable for a fusion application, but moderate concentrations of lanthanum can be tolerated.

Li8ZrO6 contains a large amount of Li and has a significant potential as a tritium breeder. However, few syntheses of fine-grain, high-purity Li8ZrO6 powder have been reported. In this study, a high-purity powder of Li8ZrO6 was synthesized by solid state reaction under hydrogen atmosphere combined with an effective lithium source and a suitable initial Li:Zr molar ratio. Mixed powders of Li2CO3 and ZrO2 were fired at around 630 °C in H2 for several hours and several firing cycles. The low firing temperature inhibited the vaporization of Li during the heating, so that excessive amounts of Li were not needed for the synthesis, and the Li:Zr ratio in the starting material was 10:1 (mol:mol). In this synthesis, Li2O was generated via the decomposition of Li2CO3 during firing in H2, and reacted with ZrO2 to form Li6Zr2O7, which reacted with itself to form Li8ZrO6.

The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

We have performed the tritium release experiment on the fusion reactor blanket at JAEA/FNS since 2009, and then clarified the ratio of tritium release and the recovered tritium chemical form. In order to acquire the detailed tritium recovery performances, we have started a new blanket tritium recovery experiment with ionization chamber (IC) at JAEA/FNS. For the appropriate tritium measurement with IC, we improved the experimental container and carried out with an intense DT neutron source at JAEA/FNS. From our new experiment, the tritium recovery radioactivity from the LSC measurement corresponds with the calculation within 6%. However, it was pointed out that further improvement in the quantitative tritium measurement by IC method was needed.

Li2TiO3 with excess Li (Li2+xTiO3+y) has been developed as an advanced tritium breeder. With respect to the tritium release characteristics of the blanket, the optimum grain size after sintering was less than 5 μm. Therefore, an emulsion method was developed to fabricate pebbles with this target grain size. The predominant factor affecting grain growth was assumed to be the presence of binder in the gel particles; this remaining binder was hypothesized to react with the excess Li, thereby generating Li2CO3, which promotes grain growth. To inhibit the generation of Li2CO3, calcined Li2+xTiO3+y pebbles were sintered under vacuum and subsequently under a 1% H2–He atmosphere. The average grain size of the sintered Li2+xTiO3+y pebbles was less than 5 μm. Furthermore, the tritium release properties of Li2+xTiO3+y pebbles were evaluated, and deuterium–tritium (DT) neutron irradiation experiments were performed at the Fusion Neutronics Source facility in the Japan Atomic Energy Agency. To remove the tritium produced by neutron irradiation, 1% H2–He purge gas was passed through the Li2+xTiO3+y pebbles. The Li2+xTiO3+y pebbles exhibited good tritium release properties, similar to those of Li2TiO3 pebbles. In particular, the released amount of tritiated hydrogen gas for easier tritium handling was greater than the released amount of tritiated water.

The accurate measurement of behavior of bred tritium released from a tritium breeder is indispensable to understand the behavior for a design of a tritium extraction system. The tritium collection method combined a CuO bed and water bubbles was not suitable to measure transient behavior of tritium released from Li2TiO3 during neutron irradiation because tritium released behavior was changed to be delayed due to adsorption of oxidized tritium on the CuO. Hence, the tritium collection method with hydrophobic catalyst instead of the CuO was demonstrated and succeeded the accurate release measurement of tritium from Li2TiO3. With the method, we assessed the behavior of tritium release under the various conditions since tritium should be released from Li2TiO3 as the form of HT as much as possible from the view point of the fuel cycle. Our results indicated; promotion of isotopic exchange reaction on the surface of Li2TiO3 by addition of hydrogen in sweep gas is mandatory in order to release tritium smoothly from Li2TiO3 irradiated with neutrons; the favorable sweep gas to release as the form of HT was hydrogen added inert gas; and the temperature of Li2TiO3 was the dominant parameter to control the chemical form of tritium released from the Li2TiO3.

Tritium breeding ceramic materials are placed at high temperatures for a long period in a fusion DEMO reactor. Therefore, the understanding of Li mass loss phenomena and its influence on tritium behavior are important. In this study, the pebbles of Li2TiO3 with excess Li were heated at 900 °C for 30 days in a 1000 Pa H2/Ar flow and tritium sorption and recovery experiments were carried out. Li mass loss by the heating was evaluated to be 0.7 wt%. The value of Li mass loss was almost same as that for 3 days heating at the same condition. Tritium sorption capacity for the heated pebbles at 600 °C and 900 °C were almost same as that for the pebbles as received. Tritium sorbed in the pebbles could not be recovered effectively by the 1000 Pa H2/Ar purge at room temperature and 300 °C but it could be recovered at 600 °C and 900 °C. The influence of the long-time heating on the behavior of tritium sorbed in the pebbles was not large.

We investigate isotope effects of lithium-ion diffusion in Lithium-ion conductors using ab initio computational methods in light of recently proposed novel lithium isotope separation technology based on ionic conductors. Using ab initio molecular dynamics simulations, we compute the diffusion coefficients D6(7) of 6Li+ (7Li+) in tetragonal Li7La3Zr2O12 and Li10GeP2S12 as a slow Li+-ion conductor and a fast Li+-ion conductor, respectively. We note that careful uncertainty estimations in the molecular dynamics simulations would be necessary to extract many-body effects on the diffusion coefficient ratio D6/D7 beyond the canonical value given by the transition-state theory. Numerical results for Li10GeP2S12 indicate that the isotope effect depends on the migration path. Specifically, the faster diffusion path along the c−axis exhibits a stronger isotope effect that is also temperature dependent, whereas the diffusion on the ab−plane shows a relatively weaker and temperature-independent effect. We introduce the probability distribution of the ionic displacement to reveal distinct differences in the ion-diffusion mechanism.

Demonstration power plant (DEMO) reactors require advanced tritium breeders with high thermal stability. Li 2 TiO 3 with excess Li (Li 2+ x TiO 3+ y ) has been developed as a material for an advanced tritium breeder. Considering the tritium release characteristics and the packing factor of the blanket, the optimum pebble diameter and grain size after sintering were determined to be 1 mm and 2+ x TiO 3+ y and Li 2 CO 3 is generated, which promotes grain growth. To inhibit the generation of Li 2 CO 3 , calcined Li 2+ x TiO 3+ y pebbles were sintered under vacuum and subsequent 1% H 2 –He atmosphere conditions. The diameter of the sintered Li 2+ x TiO 3+ y pebbles was 1.07 mm, and the average grain size on the surfaces and cross sections was 2+ x TiO 3+ y pebbles with optimum characteristics were successfully fabricated using the emulsion method.

In this study, weight reduction of Li 2 TiO 3 with excess Li and Li 4 SiO 4 at elevated temperatures under hydrogen atmosphere or water vapor atmosphere was investigated. The Li mass loss for the Li 2 TiO 3 at 900 °C was 0.4 wt% under 1000 Pa H 2 atmosphere and 1.5 wt% under 50 Pa H 2 O atmosphere. The Li mass loss for the Li 2 TiO 3 increased proportionally to the water vapor pressure in the range from 50 to 200 Pa at 900 °C and increased with increasing temperature from 700 to 900 °C although Li mass loss at 600 °C was significantly smaller than expected. It was found that water vapor concentration dependence and temperature dependence of Li mass loss for the Li 2 TiO 3 and the Li 4 SiO 4 used in this work were quite different. Water vapor is released from the ceramic breeder materials into the purge gas due to desorption of adsorbed water and water formation reaction. The released water vapor possibly promotes Li mass loss with the formation of LiOH on the surface.

The temperature of Li 2 TiO 3 pebble breeder in a fusion DEMO blanket is assumed to be more than 1000 K. For the investigation of tritium release from a Li 2 TiO 3 pebble breeder blanket at such a high temperature, we have carried out a tritium release experiment with the DT neutron source at the JAEA-FNS. The Li 2 TiO 3 pebble (1.0–1.2 mm in diameter) of 70 g was put into a stainless steel container and installed into an assembly stratified with beryllium and Li 2 TiO 3 layers. During the DT neutron irradiation, the temperature was kept at 1073 K with wire heaters in the blanket container. Helium gas including 1% hydrogen gas (H 2 /He) mainly flowed inside the container as the purge gas. Two chemical forms, HT and HTO, of extracted tritium were separately collected during the DT neutron irradiation by using water bubblers and CuO bed. The tritium activity in the water bubbler was measured by a liquid scintillation counter. To investigate the effect of moisture in the purge gas, we also performed the same experiments with H 2 O/He gas (H 2 O content: 1%) or pure helium gas. From our experiment at 1073 K, in the case of the purge gas includes H 2 , it is indicated that the increasing tendency of HT release is similar to that of the dry H 2 /He.

Li2TiO3 advanced tritium breeders with excess Li (Li2+xTiO3+y) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Considering the tritium release characteristics, the optimum grain size after sintering is >5 μm. From the results of the optimization of granulation conditions, prototype Li2+xTiO3+y pebbles with optimum grain size (>5 μm) were successfully fabricated. The Li2+xTiO3+y pebbles exhibited good tritium release properties similar to the Li2TiO3 pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water.

In the Broader Approach (BA) framework, research and development on blanket related materials and technologies have been carried out between the EU and Japan. Those activities are implemented mainly at the Rokkasho BA site in collaboration with universities in Japan. In the R&D on SiCf/SiC composites for an advanced blanket material, mechanical properties of chemically vapor infiltrated (CVI) SiCf/SiC composites have been obtained in high temperature vacuum environment up to 1000 °C. As to R&D on the tritium technology, tritium retention of the fine-grained re-crystallized tungsten has been evaluated. On reduced activation of ferritic/martensitic (RAFM) steels as the blanket structural material, a 20-ton heat of the F82H RAFM steel has been successfully produced in an electric arc furnace. Advanced neutron multiplier pebbles of beryllide (beryllium–titanium alloy) have been fabricated using a dedicated rotating electrode apparatus followed by annealing from the beryllide rod produced with the plasma sintering method. Also advanced tritium breeder pebbles of lithium-rich lithium titanate have been fabricated by an emulsion method, where the grain size, confirmed by SEM, was less than 5 μm. Tritium recovery tests of advanced tritium breeder pebbles will be investigated under 14 MeV neutron irradiation as one of the new activities.

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li2TiO3 pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

The Li2TiO3 is one of candidate breeding materials of a water cooled ceramic breeding (WCCB) blanket. In order to clarify the tritium recovery property of the WCCB blanket with the Li2TiO3 breeding material, we have performed the tritium recovery online experiment with the DT neutron source at the Fusion Neutronics Source facility in Japan Atomic Energy Agency (JAEA-FNS). We irradiated an experimental assembly simulating the WCCB blanket and recovered the tritium recovered from the Li2TiO3 pebbles put into the assembly with a heater system, sweep gases and bubblers. The activity of the recovered tritium was measured with a liquid scintillation counter. From our tritium recovery online experiment and calculation, the followings were found out: (1) the recovered tritium corresponded to the calculated tritium production within the experimental error in the range of 573–1073 K and (2) the recovered HTO tended to be easily recovered at lower temperature and high water moisture. The recovered HT increases at higher temperature and dry hydrogen circumstance. However, the maximum level of the tritium gas recovery is around 90% even at higher temperature and 1% H2 circumstance.

The vaporization property and the crystal structure of lithium metatitanate with excess Li, which has been developed as an advanced tritium breeder, were studied. After synthesizing the lithium metatitanate specimens with Li/Ti = 2.0–2.3 (at mixing of the starting materials), the vaporization properties were investigated by measuring the mass losses during heating at 1173 K. The Li excessive specimens indicated higher rates of mass loss than the stoichiometric one during heating as long as excessive lithium atom exists. The crystal structures of stoichiometric and nonstoichiometric lithium metatitanates were discussed with the result of neutron diffraction and the refined structural parameters by Rietveld analysis. The refinement ( R wp = 7.98, R p = 5.96 and R e = 2.58) suggested that lithium metatitanate with excess Li has β-Li 2 TiO 3 structure and excess Li atoms exist at unstable sites with the formation of reduced titanium or site vacancies.

The amount of 6Li in a tritium breeder decreases as a result of it being burned during the operation of a fusion reactor; however, the used tritium breeder in the replaced blanket includes a large amount of unburned 6Li. In this study, appropriate conditions and a method for mass reprocessing and reusing were evaluated. In the case of a hydrogen peroxide (H2O2) solvent, most of the Li2TiO3 was soluble at room temperature, and the solubility of Li was 96.2%. Lithium was dissolved as lithium hydroxide (LiOH⋅H2O) in an H2O2 solvent. Then a preliminary test of the preparation of raw material for Li2TiO3 pebble fabrication was carried out by using the LiOH⋅H2O dissolution liquid. The dissolution liquid was heated and exposed to water vapor in another method. The recovery rate of LiOH⋅H2O was about 100%.

Tritium fuel for fusion reactors is produced by reacting lithium-6 ( 6 Li) with neutrons in tritium breeders. This study demonstrates a method for Li recovery from seawater, wherein Li does not permeate from the anode side to the cathode side through an ionic liquid N,N,N-trimethyl-N-propylammonium–bis(trifluoromethanesulfonyl) imide. Almost all Li ions remain on the anode side (seawater), whereas the other ions in the seawater permeate to the cathode side through the ionic liquid with an applied electric voltage of 2–3 V.

Worldwide research and development has been performed on solid breeder materials as a tritium source for future fusion power reactors. These ceramic materials will be used in the International Thermonuclear Experimental Reactor (ITER) to check their performance and screen their applicability for use in fusion blankets. While all members of the ITER project are engaged in the development of solid breeder blanket concepts, five are actively involved in the fabrication of tritium breeding ceramics for the different test blanket modules to be tested in ITER. All solid breeder concepts proposed for ITER are based on the use of ceramics in the form of pebble beds with pebble diameters of about 1 mm or smaller. Although alternative materials are still under investigation, the main candidates for tritium breeding ceramics are lithium metatitanate and lithium orthosilicate. Recent developments and preparatory work for ITER are presented.

Tritium sorption capacity is an important parameter to evaluate tritium behavior on lithium ceramic breeder materials. In the present study, sorption and desorption behavior of tritiated water on Li2TiO3 with additional Li, which is in a developmental stage in Japan Atomic Energy Agency as an advanced tritium breeder materials, was observed at 20, 300, 600, and 900 °C. Tritium sorption capacity on Li2TiO3 with additional Li is larger than that on Li2TiO3. At 600 and 900 °C, the sorption capacity approximately agrees with the sum of physical adsorption capacity and chemical adsorption capacity, but at 20 and 300 °C it is smaller than that. The overall mass transfer coefficient for tritium sorption increases with temperature in the range from 20 to 600 °C but it decreases considerably at 900 °C. The sorption capacity and the mass transfer coefficient at 600 °C for the sample once used in sorption and desorption experiment at 900 °C are smaller than that for original ones.

Mixtures of tritium breeder (lithium) and neutron multiplier (beryllium) are being considered for use in increasing the tritium breeding ratio in breeding blankets. However, lithium and beryllium react under normal operating conditions, and therefore, a high-functional tritium breeder such as lithium beryllium oxide (Li2Be2O3) needs to be developed to compensate for this reaction under high-temperatures. LiOH·H2O and BeO powders were mixed in stoichiometric proportions at a Li/Be molecular ratio of 1.0. The sintering temperature was established as 1073 K by thermogravimetric/differential thermal analysis (TG–DTA). The Li/Be molar ratio of the reaction products measured by inductively coupled plasma atomic emission spectroscopy (ICP-AES) after the reaction agreed with the nominal molar ratio obtained by mixing LiOH·H2O and BeO. Crystal structure analysis of this powder was performed by the XRD technique. The XRD patterns of products were the same as those of Li2Be2O3 as listed in the JC-PDF-Card, and no impurities were indicated. The results indicate that the solid-state reaction of LiOH·H2O and BeO is suitable for synthesizing lithium beryllium oxide (Li2Be2O3).

It has been known that water vapor is released from ceramic breeder materials into the purge gas due to desorption of adsorbed water under dry atmosphere and due to the water formation reaction under hydrogen atmosphere. However, an effect of water vapor in the purge gas to Li mass loss has not been understood. In this study, mass loss of Li2TiO3 (NFI) and Li4SiO4 (FzK) under hydrogen atmosphere (1000 Pa H2/Ar), and mass loss of Li2TiO3 (NFI) and Li2TiO3 with additional Li which is in a developmental stage (JAEA) under water vapor atmosphere (50 Pa H2O/Ar) were compared, respectively. It was found that under hydrogen atmosphere Li mass loss of Li4SiO4 and Li2TiO3 is same degree although the amount of water vapor released from Li4SiO4 is larger than that from Li2TiO3. It was clarified with regard to Li2TiO3 that Li mass loss in water vapor atmosphere is larger than that in hydrogen atmosphere. Mass loss of Li2TiO3 with additional Li (JAEA) was larger than that of Li2TiO3 (NFI). It was observed by X-ray analysis that Li deposits formed on the inner wall of the quartz tube contain Li2SiO3.

Lithium (Li) procurement is becoming a matter of importance worldwide. I propose a new method for recovering Li from seawater by electrodialysis, wherein Li selectively permeates from the anode side to the cathode side through an organic membrane impregnated with an ionic liquid (PP13-TFSI). Measurements of the ion concentration at the cathode side as a function of dialysis duration showed that the Li concentration increased to 5.94% after 2 h with an applied electric voltage of 2 V. Moreover, the other ions in the seawater did not permeate the membrane. With both ends of the impregnated membrane covered with a Nafion 324 overcoat to prevent outflow of the ionic liquid, the Li concentration increased to 22.2%. This new recovery method shows good energy efficiency and is easily scalable and thus is suitable for use in seawater desalination plants.

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li 2 TiO 3 ) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.

The release behavior of tritium generated by neutron irradiation is strongly affected by the surface water of ceramic breeder materials. In the present study, the release behavior of water vapor from Li2TiO3 with added Li, which is in a developmental stage in JAEA (Japan Atomic Energy Agency) as an advanced tritium breeder materials, was observed at elevated temperatures. The capacity of chemical adsorbed water in nitrogen atmosphere and the water formation capacity in hydrogen atmosphere were quantified, respectively. The mass loss of the Li2TiO3 due to evaporation of Li-containing species at 900 °C in 10,000 ppmH2/N2 was estimated to be 1.8% of initial sample weight. The majority of evaporated Li adhered to a quartz tube around the sample bed. It was shown experimentally that the mass loss in water vapor atmosphere is larger than that in hydrogen atmosphere.

Demonstration power reactors (DEMOs) require advanced tritium breeders and neutron multipliers that have high stability at high temperatures. Lithium titanate (Li2TiO3) is one of the most promising candidates among tritium breeders because of its tritium release characteristics. Li2TiO3 with additional Li (Li2+xTiO3+y) has increased stability in a reducing atmosphere at high temperatures. In this work, Li2+xTiO3+y pebbles were fabricated using the emulsion method, which is a sol–gel method. The raw material for the fabrication of Li2+xTiO3+y pebbles was synthesized from a mixture of LiOH·H2O and H2TiO3 at specific ratios. The average diameter and the sphericity of the pebbles fabricated by the emulsion method were 1.40 mm and 1.02, respectively. In addition, beryllium (Be) intermetallic compounds (beryllides) are promising materials for advanced neutron multipliers. The results of the trial fabrications in this work showed that beryllides of Be–Ti and Be–V intermetallics could be synthesized using the plasma sintering method.

Crystal grain growth and crystallization of lithium metatitanate with excess Li (Li 2+x TiO 3+y ), which is expected as an advanced ceramic breeder for a future DEMO fusion reactor, were studied in this paper. By the observation of sintered pellets using scanning electron microscope, it was shown that the Li 2.1 TiO 3+y specimens, which have small grains (1–2 μm) and narrow size distribution, can be obtained by sintering in the temperature range from 1000 to 1100 °C. The observation also showed the rapid grain growth in the Li 2.1 TiO 3 specimens sintered above 1150 °C. The changes in the crystal structure and the phase transformation with the increase of sintering temperature were also investigated by means of powder X-ray diffraction. Finally, the relationship between the grain growth and the phase state at high temperature was discussed.

Several technical R&D activities mainly related to the blanket materials are newly launched as a part of the Broader Approach (BA) activities, which was initiated by the EU and Japan. According to the common interests for these parties in DEMO, R&Ds on reduced activation ferritic/martensitic (RAFM) steels as structural material, SiCf/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology are carried out through the BA DEMO R&D program, in order to establish the technical bases on the blanket materials and the tritium technology required for DEMO design. This paper describes overall schedule of those R&D activities and recent progress in Japan carried out by JAEA as the domestic implementing agency on BA, collaborating with Japanese universities and other research institutes.

The tritium needed as a fuel for fusion reactors is produced by the neutron capture reaction of lithium-6 ( 6 Li) in tritium breeding materials. New lithium isotope separation technique using ionic-liquid impregnated organic membranes (Ionic-Liquid-i-OMs) have been developed. Lithium ions are able to move by electrodialysis through certain Ionic-Liquid-i-OMs between the cathode and the anode in lithium solutions. In this report, the effects of protection cover and membrane thickness on the durability of membrane and the efficiency of isotope separation were evaluated. In order to improve the durability of the Ionic-Liquid-i-OM, we developed highly-durable Ionic-Liquid-i-OM. Both surfaces of the Ionic-Liquid-i-OM were covered by a nafion 324 overcoat or a cation exchange membrane (SELEMION™ CMD) to prevent the outflow of the ionic liquid. It was observed that the durability of the Ionic-Liquid-i-OM was improved by a nafion 324 overcoat. On the other hand, the organic membrane selected was 1, 2 or 3 mm highly-porous Teflon film, in order to efficiently impregnate the ionic liquid. The 6 Li isotope separation factor by electrodialysis using highly-porous Teflon film of 3 mm thickness was larger than using that of 1 or 2 mm thickness.

The crystal structure at high temperature and the thermal expansion properties of Li 2 TiO 3 , important for use as a tritium breeding material in fusion reactors, have been investigated with high resolution by using high temperature synchrotron X-ray diffraction. The temperature dependence of lattice constants and molar volumes were calculated from the data, and variation of thermal expansion around 450–500 °C has been more clearly observed than from high temperature X-ray diffraction using Cu Kα radiation. The variation of thermal expansion can be attributed to existence of a higher order phase transition or appearance of nonlinear thermal expansion.

The tritium needed as a fuel for fusion reactors is produced by the neutron capture reaction of lithium-6 ( 6 Li) in tritium breeding materials. However, natural Li contains only about 7.6 at.% 6 Li. In this paper, a new lithium isotope separation technique using an ionic-liquid impregnated organic membrane is proposed. In order to separate and concentrate lithium isotopes, only lithium ions are able to move through the membrane by electrodialysis between the cathode and the anode in lithium solutions. Preliminary experiments of lithium isotope separation were conducted using this phenomenon. Organic membranes impregnated with TMPA-TFSI and PP13-TFSI as ionic liquids were prepared, and the relationship between the 6 Li separation coefficient and the applied electrodialytic conditions was evaluated using them. The results showed that the 6 Li isotope separation coefficient in this method (about 1.1–1.4) was larger than that in the mercury amalgam method (about 1.06).

A non-stoichiometric solid state compounds of lithium titanate of the Li 2 SnO 3 -type structure with the space group C 2/ c (No. 15), Z = 8 (denoted as Li 2 TiO 3 ss) were successfully synthesized using the powders of LiOH·H 2 O and H 2 TiO 3 as the starting materials. The crystal structure of the compounds was analyzed by X-ray diffractometry using Cu Kα radiation. The Li/Ti ratio in the compounds was analyzed by an inductively coupled plasma–optical emission spectroscopy. The microstructure was analyzed by scanning electron microscopy. Thermal diffusivity was measured by the laser flash method. The influences of the Li/Ti ratio on microstructure, Li density, and thermal diffusivity were investigated. Li density of the compounds increased due to the synergy effect of the good sintering behavior and the Li-rich composition. The increase in the Li/Ti ratios of the compounds encouraged the sintering and increased the thermal diffusivity.

Lithium titanate (Li 2 TiO 3 ) is one of the most promising candidates among tritium breeding materials because of its good tritium release characteristics. However, the mass of Li 2 TiO 3 decreased with time in a hydrogen atmosphere by the reduction of Ti and Li evaporation. In order to prevent the mass decrease at high temperatures, advanced tritium breeding material with added Li (Li 2+ x TiO 3+ y ) should be developed. For this purpose, an advanced Li 2 TiO 3 with added Li was synthesized from proportionally mixed LiOH·H 2 O and H 2 TiO 3 with a Li/Ti ratio of 2.2. The results of X-ray diffraction measurement showed that this advanced tritium breeding material existed as the non-stoichiometric compound Li 2+ x TiO 3+ y . The desired molar ratio of Li/Ti was achieved by appropriate mixing of LiOH·H 2 O and H 2 TiO 3 . Therefore, synthesis by mixing LiOH·H 2 O and H 2 TiO 3 is a promising mass production method for the advanced tritium breeding material with added Li for the test blanket module of ITER.

Lithium titanate (Li2TiO3) is one of the most promising candidates among tritium breeding materials because of its good tritium release characteristics. However, the mass of Li2TiO3 decreased with time by Li evaporation in a hydrogen atmosphere and by Li burn-up. In order to prevent the mass decrease at high temperatures, Li2TiO3 with added Li (Li2+xTiO3+y) have been developed as one of advanced tritium breeders. Li2TiO3 with added Li was synthesized from proportionally mixed LiOH·H2O and H2TiO3 with a Li/Ti ratio of 2.2, and then the sol–gel method were selected as the candidate for the pebble fabrication technique of advanced tritium breeders. In this study, trial fabrication of pebbles was performed by the sol–gel method. The diameter of the pebbles is 1.18 mm, and the sphericity is 1.04. The density of the pebbles is about 89%T.D analyzed by mercury intrusion technique. Molar ratio (Li/Ti) of the sintered pebbles evaluated by inductively coupled plasma atomic emission spectrometry (ICP-AES) is 2.12. These results show that the pebble fabrication using the sol–gel method is a promising production technique for mass production of the advanced tritium breeder pebbles.

Lithium titanate (Li 2 TiO 3 ) is one of the most promising candidates among the proposed solid breeder materials for fusion reactors. Addition of H 2 to inert sweep gas has been proposed for enhancing the release of bred tritium from breeder material. However, the mass of Li 2 TiO 3 has been found to decrease with time in the hydrogen atmosphere. This mass change indicates that the oxygen content of the sample decreased, suggesting the change from Ti 4+ to Ti 3+ . Development of Li 2 TiO 3 with Li 4 TiO 4 additive is expected to be effective in control the mass change at the time of high temperature use. In the present paper, synthetic methods of advance lithium titanate (Li 2 TiO 3 added with Li 4 TiO 4 ) have been extensively investigated by solid state reaction. The crystal structure of the sample was analyzed by means of inductively coupled plasma atomic emission spectrometry (ICP-AES) and X-ray diffraction (XRD). The molar ratios Li 2 O/TiO 2 of samples after the reaction were in good agreement with the values at the time of mixing LiOH·H 2 O and H 2 TiO 3 . The overall results suggest that the solid state reaction of LiOH·H 2 O and H 2 TiO 3 is one of the most appropriate synthesis methods for advanced lithium titanate with Li 4 TiO 4 in prospect of mass production technology for the test blanket module (TBM) of International Thermonuclear Experimental Reactor (ITER).

Li 2 TiO 3 is one of the most promising candidates among solid breeder materials proposed for fusion reactors. However, the mass of Li 2 TiO 3 was found to decrease with time in the sweep gas mixed with hydrogen. This mass change indicates that the oxygen content of the sample decreased, suggesting the change from Ti 4+ to Ti 3+ . In the present paper, the crystal structure and the non-stoichiometry of Li 2 TiO 3 added with Li 2 O have been extensively investigated by means of X-ray diffraction (XRD) and thermogravimetry. In the case of the Li 2 TiO 3 samples used in the present study, LiO–C 2 H 5 or LiO–i–C 3 H 7 and Ti(O–i–C 3 H 7 ) 4 were mixed in the proportion corresponding to the molar ratio Li 2 O/TiO 2 of either 2.00 or 1.00. In thermogravimetry, the mass of this sample decreased with time due to lithium deficiency, where no presence of oxygen deficiency was indicated.

In order to develop the reprocessing technology of lithium ceramics (Li 2 TiO 3 , CaO-doped Li 2 TiO 3 , Li 4 SiO 4 and Li 2 O) as tritium breeder materials for fusion reactors, the dissolution methods of lithium ceramics to recover 6 Li resource and the purification method of their lithium solutions to remove irradiated impurities ( 60 Co) were investigated. In the present work, the dissolving rates of lithium from each lithium ceramic powder using chemical aqueous reagents such as HN0 3 , H 2 O 2 and citric acid (C 6 H 8 O 7 · H 2 O) were higher than 90%. Further the decontamination rate of 60 Co added into the solutions dissolving lithium ceramics was higher than 97% using the activated carbon impregnated with 8-hydroxyquinolinol as chelate agent.