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Journal Articles

Pebble fabrication of super advanced tritium breeders using a solid solution of Li $_{2+x}$ TiO $_{3+y}$ with Li $_{2}$ ZrO $_{3}$

Hoshino, Tsuyoshi

Nuclear Materials and Energy (Internet), 9, p.221 - 226, 2016/12

https://doi.org/10.1016/j.nme.2016.05.004

Demonstration power plant (DEMO) reactors require advanced tritium breeders with high tritium breeding ratios (TBRs) and increased stability at high temperatures. Li $_{2}$ TiO $_{3}$ with excess Li (Li $_{2+x}$ TiO $_{3+y}$ ) is one of the candidate materials for advanced tritium breeders. While Li $_{2+x}$ TiO $_{3+y}$ pebbles have some good properties, the TBR in Li $_{2+x}$ TiO $_{3+y}$ is smaller than that in Li $_{2}$ O because of its low Li density. Therefore, new Li-contained ceramic composites with not only high stability but also high Li density have been developed in JAEA. I focused on a solid solution with a new characteristic. Since the atomic radius of Li and Zr is similar, I surveyed the possibility of solid solution for a Li $_{2}$ O-TiO $_{2}$ -ZrO $_{2}$ (LTZO) system. The solid solution pebbles of LTZO are fabricated by an emulsion method are designated as LTZO20 (Li $_{2+x}$ TiO $_{3+y}$ with added 20 wt% Li $_{2}$ ZrO $_{3}$ ). The X-ray diffraction patterns of the sintered LTZO20 pebbles are approximately the same as those of Li $_{2+x}$ TiO $_{3+y}$ pebbles, and then Li $_{2}$ ZrO $_{3}$ disappeared. The results show that LTZO pebbles are not a two-phase material but a solid solution.

Journal Articles

Evaluation of tritium release behavior from Li $_{2}$ TiO $_{3}$ during DT neutron irradiation by use of an improved tritium collection method

Edao, Yuki; Kawamura, Yoshinori; Hoshino, Tsuyoshi; Ochiai, Kentaro

Fusion Engineering and Design, 112, p.480 - 485, 2016/11

https://doi.org/10.1016/j.fusengdes.2016.06.013

Times Cited Count：8 Percentile：54.33(Nuclear Science & Technology)

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 Li $_{2}$ TiO $_{3}$ 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 Li $_{2}$ TiO $_{3}$ . With the method, we assessed the behavior of tritium release under the various conditions since tritium should be released from Li $_{2}$ TiO $_{3}$ 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 Li $_{2}$ TiO $_{3}$ by addition of hydrogen in sweep gas is mandatory in order to release tritium smoothly from Li $_{2}$ TiO $_{3}$ irradiated with neutrons; the favorable sweep gas to release as the form of HT was hydrogen added inert gas; and the temperature of Li $_{2}$ TiO $_{3}$ was the dominant parameter to control the chemical form of tritium released from the Li $_{2}$ TiO $_{3}$ .

Journal Articles

A New blanket tritium recovery experiment with intense DT neutron source at JAEA/FNS

Ochiai, Kentaro; Edao, Yuki; Hoshino, Tsuyoshi; Kawamura, Yoshinori; Ota, Masayuki; Kwon, Saerom; Konno, Chikara

Fusion Engineering and Design, 109-111(Part B), p.1143 - 1147, 2016/11

https://doi.org/10.1016/j.fusengdes.2016.01.002

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 more detail 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 the further improvement on the quantitative tritium measurement by IC method was needed.

Journal Articles

Pebble fabrication and tritium release properties of an advanced tritium breeder

Hoshino, Tsuyoshi; Edao, Yuki; Kawamura, Yoshinori; Ochiai, Kentaro

Fusion Engineering and Design, 109-111(Part B), p.1114 - 1118, 2016/11

https://doi.org/10.1016/j.fusengdes.2016.01.012

Li $_{2}$ TiO $_{3}$ with excess Li (Li $_{2+x}$ TiO $_{3+y}$ ) has been developed as an advanced tritium breeder. Considering the tritium release characteristics, the optimum grain size of pebble is less than 5 m. Therefore, the pebble fabrication by using emulsion method was carried out to obtain the target value. Calcined Li $_{2+x}$ TiO $_{3+y}$ pebbles were sintered under vacuum and subsequent 1% H $_{2}$ -He atmosphere. The average grain size of the sintered pebbles was less than 5 m. Furthermore, the tritium release properties of the pebbles are required for DEMO blanket design. In the present study, an evaluation of the tritium release properties of the pebbles was performed by DT neutron irradiation. The Li $_{2+x}$ TiO $_{3+y}$ pebbles exhibited good tritium release properties similar to the Li $_{2}$ TiO $_{3}$ pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water.

Journal Articles

Progress of R&D on water cooled ceramic breeder for ITER test blanket system and DEMO

Kawamura, Yoshinori; Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Edao, Yuki; Hayashi, Takumi; Hoshino, Tsuyoshi; et al.

Fusion Engineering and Design, 109-111(Part B), p.1637 - 1643, 2016/11

https://doi.org/10.1016/j.fusengdes.2015.11.002

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 design activities. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

Journal Articles

Experimental investigation on tritium release from lithium titanate pebble under high temperature of 1073 K

Ochiai, Kentaro; Edao, Yuki; Kawamura, Yoshinori; Hoshino, Tsuyoshi; Ota, Masayuki; Sato, Satoshi; Konno, Chikara

Fusion Engineering and Design, 98-99, p.1843 - 1846, 2015/10

https://doi.org/10.1016/j.fusengdes.2015.07.003

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 recovery from Li $_{2}$ TiO $_{3}$ pebble breeder blanket at such a high temperature, we have carried out a tritium recovery experiment with the DT neutron source at the JAEA-FNS. The Li $_{2}$ TiO $_{3}$ pebble 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, respectively. Helium gas including 1% hydrogen gas (H $_{2}$ /He) was mainly flowed inside the container as the sweep 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 sweep 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, it is shown that the tritium recovery corresponded with the calculation with MCNP code. It was also indicated that the ratio of recovered HT with temperature gas tended to be higher.

Journal Articles

Optimization of sintering conditions of advanced tritium breeder pebbles fabricated by the emulsion method

Hoshino, Tsuyoshi

Fusion Engineering and Design, 98-99, p.1788 - 1791, 2015/10

https://doi.org/10.1016/j.fusengdes.2015.04.023

Times Cited Count：26 Percentile：87.30(Nuclear Science & Technology)

Demonstration power reactors require advanced tritium breeders that have high stability at high temperatures. Therefore, an original material of Li $_{2}$ TiO $_{3}$ with excess Li (Li $_{2+x}$ TiO $_{3+y}$ ) as an advanced tritium breeder that can make up to the lithium loss has been proposed. Furthermore, The author has been developing a technique of fabricating Li $_{2+x}$ TiO $_{3+y}$ pebbles using the emulsion method. Considering the tritium release characteristics and the packing factor of the blanket, the desired pebble diameter and grain size after sintering were 1 mm and 5 m, respectively. To prevent the grain growth of pebbles, calcined Li $_{2+x}$ TiO $_{3+y}$ pebbles were sintered in a vacuum atmosphere at 1073 K for 3 h and in a 1%H $_{2}$ -He atmoshere at 1323K for 5 h. The average grain size on the surfaces and cross sections of the sintered Li $_{2+x}$ TiO $_{3+y}$ pebbles was 5 m. In addition, the diameter of sintered Li $_{2+x}$ TiO $_{3+y}$ pebbles was 1.07 mm.

Journal Articles

Innovative lithium recovery technique from seawater by using world-first dialysis with a lithium ionic superconductor

Hoshino, Tsuyoshi

Desalination, 359, p.59 - 63, 2015/03

https://doi.org/10.1016/j.desal.2014.12.018

Times Cited Count：143 Percentile：97.78(Engineering, Chemical)

Lithium (Li) procurement is becoming a matter of importance worldwide. Herein, I propose a method for recovering Li from seawater by using world-first dialysis, wherein Li only permeates from the negative electrode side to the positive electrode side through a Li ionic superconductor functioning as a Li separation membrane (LISM). Measurements of the Li ion concentration at the positive electrode side as a function of dialysis duration showed that the Li concentration increased to approximately 7% after 72 h with no applied electric voltage. Moreover, other ions in the seawater did not permeate the LISM. With both ends of the LISM bound with a negative and positive electrode, hydrated Li ion was transformed to Li ion only because Li ion can permeate through the LISM. This new recovery method shows good energy efficiency and is easily scalable and is thus suitable for the industrialized mass production of Li in South American countries, which have briny water containing Li.

Journal Articles

Evaluation of tritium release properties of advanced tritium breeders

Hoshino, Tsuyoshi; Ochiai, Kentaro; Edao, Yuki; Kawamura, Yoshinori

Fusion Science and Technology, 67(2), p.386 - 389, 2015/03

https://doi.org/10.13182/FST14-T35

Times Cited Count：15 Percentile：72.86(Nuclear Science & Technology)

Demonstration power reactors (DEMOs) require advanced tritium breeders that have high stability at high temperatures. Therefore, the pebble fabrication of Li $_{2}$ TiO $_{3}$ with excess Li (Li $_{2+x}$ TiO $_{3+y}$ ) as an advanced tritium breeder was carried out. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed. DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. The Li $_{2+x}$ TiO $_{3+y}$ pebbles exhibited good tritium release properties similar to the Li $_{2}$ TiO $_{3}$ pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water.

Journal Articles

Trial examination of direct pebble fabrication for advanced tritium breeders by the emulsion method

Hoshino, Tsuyoshi

Fusion Engineering and Design, 89(7-8), p.1431 - 1435, 2014/10

https://doi.org/10.1016/j.fusengdes.2013.12.052

Times Cited Count：22 Percentile：81.74(Nuclear Science & Technology)

Demonstration power reactors (DEMOs) require advanced tritium breeders with high thermal stability. For the mass production of advanced tritium breeder pebbles, pebble fabrication by the emulsion method is a promising technique. To develop the most efficient pebble fabrication method, a new direct pebble fabrication process utilizing the emulsion method was implemented. A prior pebble fabrication process consisted of the preparation of raw materials followed by granulation. The new process integrates the preparation and granulation of raw materials. The slurry for the emulsion granulation of Li $_{2}$ TiO $_{3}$ as a tritium breeder consists of mixtures of Li $_{2}$ CO $_{3}$ and TiO $_{2}$ at specific ratios. The average diameter of the obtained sintered Li $_{2}$ TiO $_{3}$ pebbles were 1.0 mm. The trial fabrication results suggest that the new process has the potential to increase the fabrication efficiency of advanced tritium breeder pebbles.

Journal Articles

DT neutron irradiation experiment for evaluation of tritium recovery from WCCB blanket

Ochiai, Kentaro; Kawamura, Yoshinori; Hoshino, Tsuyoshi; Edao, Yuki; Takakura, Kosuke; Ota, Masayuki; Sato, Satoshi; Konno, Chikara

Fusion Engineering and Design, 89(7-8), p.1464 - 1468, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.04.079

Times Cited Count：9 Percentile：52.93(Nuclear Science & Technology)

We have performed the tritium recovery experiment on fusion reactor blanket with DT neutrons at the Fusion Neutronics Source facility in Japan Atomic Energy Agency. The candidate breeding material, Li $_{2}$ TiO $_{3}$ pebble, was put into the container which was set up it into an assembly simulating water cooled ceramic breeding (WCCB) blanket. Helium sweep gas including H $_{2}$ (1%) and/or H $_{2}$ O (1%) was flowed and extracted tritium was collected to water bubblers during DT neutron irradiation. The Li $_{2}$ TiO $_{3}$ pebble was also heated up to a constant temperature at 573, 873 and 1073 K, respectively. We arranged the tritium recovery system to measure tritiated water moisture and tritium gas, separately, and to investigate the amount of recovered tritium and the chemical form. From our experiments, it was showed that the amount of recovered tritium was corresponded to the calculation value and the ratio of chemical form depended to the temperature and kinds of sweep gas.

Journal Articles

Blanket material and technology developments toward DEMO under the Broader Approach framework

Nishitani, Takeo; Yamanishi, Toshihiko; Tanigawa, Hiroyasu; Nakamichi, Masaru; Nozawa, Takashi; Hoshino, Tsuyoshi; Ochiai, Kentaro

Fusion Engineering and Design, 89(7-8), p.1699 - 1703, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.01.075

On the Broader Approach framework, R&D on the 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 Japan. In the R&D on SiC/SiC composites for an advanced blanket material, CVI-SiC/SiC composites have been obtained in high temperature vacuum environment up to 1000 $^{circ}$ C. As the R&D on the tritium technology, tritium retention of the fine-grained re-crystallized tungsten has been evaluated. On reduced activation ferritic/martensitic (RAFM) steels as the blanket structural material, 20-ton heat of the F82H RAFM steel has been successfully conducted by an electric arc furnace. Advanced neutron multiplier pebbles of beryllide have been fabricated with dedicated rotating electrode apparatus followed by annealing. Also advance tritium breeder has been fabricated by an emulsion method, where the grain size is confirmed by the SEM to be smaller than 5 -m.

Journal Articles

R&D status on water cooled ceramic breeder blanket technology

Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.

Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.01.035

Times Cited Count：22 Percentile：81.74(Nuclear Science & Technology)

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. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed 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 simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.

Journal Articles

Development of technology for recovering lithium from seawater by electrodialysis using ionic liquid membrane

Hoshino, Tsuyoshi

Fusion Engineering and Design, 88(11), p.2956 - 2959, 2013/11

https://doi.org/10.1016/j.fusengdes.2013.06.009

Times Cited Count：75 Percentile：98.13(Nuclear Science & Technology)

The tritium as a fuel for fusion reactors is produced by the reaction of lithium-6 ( $^{6}$ Li) with neutron in tritium breeding material. However, Li is one of 31 rare metal elements. Furthermore, as a means of addressing global warming, the world is increasingly turning to the use of Li-ion batteries in electric vehicles and as storage batteries in the home; therefore, there is a growing need for Li. In view of Japanese high dependence on imports for material resources, the securing of enough Li resources is an important policy for domestic industry in Japan. We proposed new method for Li recovery from seawater. The method involves the use of an ionic liquid through which only the Li ions in seawater and not the other ions, including Na, Mg, Ca and K, permeate from the anode side to the cathode side during electrodialysis. Thus, the Li ions become concentrated on the cathode side and can be recovered. With both ends of an ionic liquid covered with a SELEMION $^{TM}$ to prevent outflow of the ionic liquid, Li concentration increased from 4.5% after 2 h to 11.0% after 24 h with an applied electric voltage of 2 V.

Journal Articles

Development of methods for reprocessing and reuse of tritium breeder materials in broader approach activities

Hoshino, Tsuyoshi

Journal of Nuclear Materials, 442(1-3), p.S425 - S428, 2013/11

https://doi.org/10.1016/j.jnucmat.2013.03.009

Times Cited Count：4 Percentile：29.90(Materials Science, Multidisciplinary)

Development of reprocessing and reuse of this spent breeder including unburned $^{6}$ Li is important from viewpoint of effective utilization $^{6}$ Li as a limited resource. This activity has been started between Japan and the EU in the Broader Approach activities. The effect of solvent on dissolution of Li $_{2}$ TiO $_{3}$ powder was evaluated. Li $_{2}$ TiO $_{3}$ powder with 99.9% purity was prepared as a starting powder. Solubility of Li was evaluated from the measurement result of the Li content in solution by ICP (inductively coupled plasma) atomic emission spectrometry. Peroxide hydrogen (H $_{2}$ O $_{2}$ ) was selected as solvent. Most of Li $_{2}$ TiO $_{3}$ was soluble in room temperature, and then the solubility of Li were larger than 90%.

Journal Articles

Development of fabrication technologies for advanced tritium breeder pebbles by the sol-gel method

Hoshino, Tsuyoshi

Fusion Engineering and Design, 88(9-10), p.2264 - 2267, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.05.025

Times Cited Count：17 Percentile：74.86(Nuclear Science & Technology)

Lithium titanate (Li $_{2}$ TiO $_{3}$ ) has been recognized as a prominent candidate material for use in a tritium breeder. However, the mass of Li $_{2}$ TiO $_{3}$ is decreased over time by Li evaporation in a hydrogen atmosphere, Li burn-up under the high temperatures, and high neutron fluence irradiation found in a DEMO reactor. To compensate for this decrease in mass at high temperatures, Li $_{2}$ TiO $_{3}$ with additional Li (Li $_{2+x}$ TiO $_{3+y}$ ) have been developed as an advanced tritium breeder. Pebble fabrication using the sol-gel method is one of the promising techniques for the mass production of the advanced tritium breeder pebbles. The authors have been developing a technique of fabricating Li $_{2+x}$ TiO $_{3+y}$ pebbles using the sol-gel method. To increase the density of sintered Li $_{2+x}$ TiO $_{3+y}$ pebbles, the sintering temperature was changed, and at 1473 K, the density of the pebbles was increased to approximately 75%T.D., without any increase in the grain size. This shows the pore size in sintered Li $_{2+x}$ TiO $_{3+y}$ pebbles is decreased by vacuum sintering.

Journal Articles

Trial synthesis of Li $_{2}$ Be $_{2}$ O $_{3}$ for high-functional tritium breeders

Hoshino, Tsuyoshi; Oikawa, Fumiaki; Natori, Yuri*; Kato, Kenichi*; Sakka, Tomoko*; Nakamura, Mutsumi*; Tatenuma, Katsuyoshi*

Fusion Engineering and Design, 88(9-10), p.2268 - 2271, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.05.013

Times Cited Count：4 Percentile：29.90(Nuclear Science & Technology)

Lithium titanate with additional Li (Li $_{2+x}$ TiO $_{3+y}$ ) and lithium orthosilicate (Li $_{4}$ SiO $_{4}$ ) is one of the most promising candidates for use in a tritium breeder because of its good chemical and mechanical stabilities. Currently, mixtures of tritium breeder pebble and neutron multiplier (Be or Be $_{12}$ Ti) pebble are being considered for use in increasing the tritium breeding ratio in a breeding blanket. However, lithium and beryllium are gradually reacted under practical operating conditions, and therefore a high-functional tritium breeder such as lithium beryllium oxide (Li $_{2}$ Be $_{2}$ O $_{3}$ ) needs to be developed to compensate for this reactive characteristic under high temperature use. In this study, methods of synthesizing Li $_{2}$ Be $_{2}$ O $_{3}$ have been extensively investigated by means of solid-phase reaction. The solid-phase reaction of LiOH(H $_{2}$ O) and BeO is a suitable synthesis method for lithium beryllium oxide (Li $_{2}$ Be $_{2}$ O $_{3}$ ). It is expected that single-phase Li $_{2}$ Be $_{2}$ O $_{3}$ will be stable under the mixture conditions of a tritium breeder and neutron multiplier in the blanket region at high temperatures.

Journal Articles

Preliminary studies of lithium recovery technology from seawater by electrodialysis using ionic liquid membrane

Hoshino, Tsuyoshi

Desalination, 317, p.11 - 16, 2013/05

https://doi.org/10.1016/j.desal.2013.02.014

Times Cited Count：141 Percentile：97.30(Engineering, Chemical)

Lithium (Li), one of the 31 rare metal elements among the 112 known elements, is fast becoming a valuable commodity. As a means of addressing global warming, the world is increasingly turning to the use of Li-ion batteries in electric vehicles and as storage batteries in the home; therefore, there is a growing need for Li. Li is primarily recovered from salt lakes in South America but is also present in seawater. Thus, 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 ionic liquid. Hoshino developed a novel Li-isotope separation technique that uses an ionic liquid. This technique could also be used to recover Li from seawater. 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 an ionic liquid. With both ends of an ionic liquid covered with a MEA (Membrane Electrode Assembly) to prevent outflow of the ionic liquid, the Li concentration increased. This new recovery method shows good energy efficiency and is easily scalable and thus is suitable for use in seawater desalination plants.

Journal Articles

A Feasible DEMO blanket concept based on water cooled solid breeder

Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Tanigawa, Hisashi; Enoeda, Mikio; Tanigawa, Hiroyasu; Nakamichi, Masaru; et al.

Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03

This paper presents the conceptual design of a blanket with simplified structure whose interior consists of the mixture of breeder and multiplier pebble bed, cooling tubes and support for them only. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in TBR even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production. On the other hand, the thickness of blanket housing is important from the viewpoint of safety. The blanket housing may rupture when the cooling pipe in the blanket is tearing, because thickness of structure materials is thin as 22 mm. This thickness is expected to maintain to 8 MPa in the steam pressure. Finally, the blanket housing, and aspect ratio of blanket shape is proposed in consideration of TBR, and engineering problem such as maintenance and manufacture are discussed.

JAEA Reports

Characteristic of tritium release from Li $_{2}$ TiO $_{3}$ under D-T neutron irradiation

Edao, Yuki; Kawamura, Yoshinori; Ochiai, Kentaro; Hoshino, Tsuyoshi; Takakura, Kosuke; Ota, Masayuki; Iwai, Yasunori; Yamanishi, Toshihiko; Konno, Chikara

JAEA-Research 2012-040, 15 Pages, 2013/02

JAEA-Research-2012-040.pdf:1.8MB

Tritium generation and recovery studies on Li $_{2}$ TiO $_{3}$ as a solid breeding material under neutron irradiation carried out in the Fusion Neutron Source (FNS) facility. A capsule with Li $_{2}$ TiO $_{3}$ packed bed was put in a system which simulated an actual blanket system which built in beryllium blocks and lithium titanate ones. Estimated values of the amount of tritium generation by a numerical calculation agreed closely with experimental values. The capsule was heated up to 300 $^{circ}$ C, and helium, helium with water vapor, hydrogen or hydrogen/water vapor were selected as purge gas. In the case of purge by helium added water vapor, the ratio of HTO to total tritium release was 98%. In helium with hydrogen/water vapor purge, the ratio of HTO to total tritium release was 80%, which was confirmed that HTO released by isotope exchange reaction between water vapor and tritium. In helium with hydrogen purge, the ratio of HT to total tritium release was 6070%, which was shown that HT released by isotope exchange reaction between hydrogen gas and tritium. HTO released by water generation reaction between hydrogen in purge gas and oxygen in Li $_{2}$ TiO $_{3}$ although water vapor was not added in purge gas. The ratio of HTO release seemed to be small under the deoxidized condition of the Li $_{2}$ TiO $_{3}$ surface. Tritium release behavior in the Li $_{2}$ TiO $_{3}$ depended on the composition of purge gas, and its chemical form was affected by the surface conditions of Li $_{2}$ TiO $_{3}$ .