Masaru Nakamichi

Application of beryllium intermetallic compounds to neutron multiplier of fusion blanket

Abstract Beryllium metal is commonest as a neutron multiplier of fusion blanket. However, the application of beryllium metal to DEMO reactor that requires higher temperature (600–900xa0°C) is very severe, because the melting point is low (1285xa0°C) and the starting temperature of swelling by neutron irradiation is low (∼500xa0°C). Therefore beryllium intermetallic compounds that have high melting point have been expected as the advanced material for fusion DEMO blanket. Several R&D concerning the characterization and the fabrication technology for beryllium intermetallic compounds have been performed in Japan that takes a leading part in these R&D. From the results of these R&D, it made clear that beryllium intermetallic compounds had excellent property and the feasibility as the neutron multiplier for fusion DEMO blanket. In addition, fabrication technology for the pebbles was studied.

Characterization of Chemical Densified Coating as Tritium Permeation Barrier

In a fusion blanket design, ceramic coating on structural materials has been considered to be used as a tritium permeation barrier. The Chemical Densified Coating (CDC) method has some advantage compared with another coating method. This method is capable to form densified coating on either the outer or the inner surface of a tube or a container. This process temperature is low (450°C). The fabrication technique of Cr2O3-SiO2 coating had been developed using CDC method. However, Cr2O3-SiO2 coating had open pores in the coating. For filling open pores, the densification treatment by CrPO4 was examined. In this study, the verification of open pores, the thermal shock resistivity, the adhesion strength and the deuterium permeability were evaluated and compared with Cr2O3-SiO2 (Type 1) coating and Cr2O3-SiO2 including CrPO4 (Type 2) coating. From these results, it was confirmed that Type 2 coating had a good adhesion property, and permeation reduction factor of SS316 with Cr2O3-SiO2 including CrPO4 coating reached about 1,000 at 600°C.

In-situ Tritium Recovery Experiments of Blanket In-pile Mockup with Li2TiO3 Pebble Bed in Japan

Lithium titanate (Li2TiO3) is one of the most attractive tritium breeders for breeding blanket in fusion reactor from view points of low tritium inventory, high chemical stability and so on. The data on the performance of a blanket mockup with pebble bed under neutron irradiation is needed for the design of breeding blanket. To obtain such data, two kinds of the blanket in-pile mockups with Li2TiO3 pebble bed were developed and the in-situ tritium recovery experiments were carried out in the Japan Materials Testing Reactor (JMTR). In these studies, effects of various parameters, i.e., irradiation temperature, sweep gas flow rate, etc., on the tritium recovery behavior from Li2TiO3 pebble bed were evaluated. It was found that the tritium recovery (R) to tritium generation (G) ratio (R/G) increased with increasing the temperature of Li2TiO3 pebble bed and was saturated when the temperature of Li2TiO3 pebble bed at the outside edge exceeded 300°C. Additionally, the sweep gas flow rate in the range of 100 to 900 cm3/min affected very little the tritium recovery from Li2TiO3 pebble bed. A good prospect for the design of breeding blankets using Li2TiO3 pebble bed was obtained from these results of in-situ experiments.

Compatibility of yttria (Y2O3) with liquid lithium

Abstract Compatibility of sintered specimens and plasma sprayed coating specimens, of Y 2 O 3 with liquid lithium was tested at 773 K. No configuration change was observed for the sintered specimens with a slight increase of thickness for 1419 h. Lithium—yttrium complex oxide (LiYO 2 ) was formed on the surface, and the inner part changed to gray or black nonstoichiometry Y 2 O 3− x with lower electrical resistivity. The plasma sprayed coating specimens were severely attacked by liquid lithium with or without applied electric field. Lithium penetrated into the coating layer through small cracks and reacted on Y 2 O 3 to form LiYO 2 , which has a different density from Y 2 O 3 and is more brittle than Y 2 O 3 . It is concluded that Y 2 O 3 has a possibility as a ceramic coating material for liquid blankets if it can be made into a dense coating on the surface of piping materials.

Development of advanced blanket materials for a solid breeder blanket of a fusion reactor

The design of an advanced solid breeding blanket in a DEMO reactor requires a tritium breeder and a neutron multiplier that can withstand high temperatures and high neutron fluences, and the development of such advanced blanket materials has been carried out by collaboration between JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by a wet process is a reference material as a tritium breeder, but its stability at high temperatures has to be improved for its application in a DEMO blanket. One of these improved materials, TiO2-doped Li2TiO3 pebbles, was successfully fabricated and studied. For the advanced neutron multiplier, beryllides that have a high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that it had lower swelling and tritium inventory than beryllium metal. Pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. These activities have shown that there is a bright prospect in realizing a DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides.

Tritium permeation through austenitic stainless steel with chemically densified coating as a tritium permeation barrier

Chemically densified coating formed on the surface of austenitic stainless steel (SUS 316) was examined for compatibility with molten lithium-lead eutectic alloy (Li17ue5f8Pb83) and tritium permeability. The chemically densified coating (CDC) consisting of SiO2 particles and a Cr2O3 matrix with a thickness of 60 μm was unstable in contact with the molten alloy as predicted from a thermodynamic calculation at 600°C, and it was degraded in several days. In an in-pile experiment, specimens with the coating on the front surface or the rear surface were immersed in Li17ue5f8Pb83 molten alloy, and their tritium permeabilities were measured. The permeability of the former was reduced to 110 of the ideal value in the diffusion-limited case, while that of the latter was less than 1100 of the diffusion-limited value even in a pure H2 atmosphere. It is concluded that CDC is quite effective to reduce tritium permeability in the condition of not contacting molten Li17ue5f8Pb83 alloy.

Integrated experiment of blanket in-pile mockup with Li2TiO3 pebbles

Abstract Lithium titanate (Li 2 TiO 3 ) is one of the candidate tritium breeding materials for breeding blanket of the fusion reactor. Therefore, tritium release experiments from Li 2 TiO 3 packing region were carried out to evaluate the effects of various parameters, i.e. sweep gas flow rate, irradiation temperature, and hydrogen content in sweep gas, etc. on tritium release. As for the shape of the Li 2 TiO 3 , a small spherical form (pebbles) is preferred to reduce the induced thermal stress in the breeding material. The diameter and total weight of Li 2 TiO 3 pebbles were 1 mm and ∼135 g, respectively. And the integrated experiment of blanket in-pile mockup with Li 2 TiO 3 pebble bed was carried out at the Japan Materials Testing Reactor (JMTR). The tritium released from Li 2 TiO 3 pebble bed was swept by the helium gas with hydrogen. The total tritium concentration (HT+HTO) and gaseous tritium concentration (HT) of tritium released from Li 2 TiO 3 pebbles were measured, and HT/(HT+HTO) ratio was evaluated under various conditions.

Material design of ceramic coating by plasma spray method

Abstract In the ceramic coating on substrate, cracking and peeling occur due to the difference of thermal expansion between substrate material and coating material. For evaluation of peeling property of plasma sprayed coating, it is demanded that thermal properties of plasma sprayed coating are estimated in detail. In this study, the results of comparison of thermal properties between bulk material and plasma sprayed material are investigated to design the ceramic coating quantitatively. Thermal conductivity of plasma sprayed MgO·Al 2 O 3 is decreased by approximately 50% to that of sintered MgO·Al 2 O 3 . Thermal conductivity of plasma sprayed 410SS agreed well with the calculation results of relation between porosity and thermal conductivity of iron sintered material. Thermal expansions of atmospheric plasma sprayed MgO·Al 2 O 3 and bulk 410SS, respectively. Therefore, as to material design on ceramic coating, it was made clear that thermal conductivity is more important than thermal expansion.

Characterization of Y2O3 coating under neutron irradiation

Abstract Ceramic coatings on the surface of structural materials such as 316SS have been considered for electrical insulators and tritium permeation barriers in fusion reactor designs. Y 2 O 3 is one of the most promising materials as a coating from a point of high electrical resistivity, etc. In this report, the electrical conductivity of a Y 2 O 3 coating was investigated under neutron irradiation with the Japan Materials Testing Reactor (JMTR). The specimen was 316SS with a Y 2 O 3 coating, and was irradiated at 300°C in a He atmosphere. From the results of in situ measurements on the electrical conductivity of a Y 2 O 3 coating, the Radiation Induced Conductivity (RIC) was observed at JMTR power-up. The electrical conductivity of the coating before neutron irradiation and under neutron irradiation were about 3xa0×xa010 −13 and about 1xa0×xa010 −9 1/Ω/cm, respectively. The electrical conductivity of the coating was constant under neutron irradiation but after neutron irradiation recovered up to that before neutron irradiation. The Radiation Induced Electrical Degradation (RIED) in the Y 2 O 3 coating was not recognized up to a fluence of about 5xa0×xa010 20 n/cm 2 (>1 MeV). It was clear that the coating had good electrical resistivity under neutron irradiation corresponding to about 4xa0×xa010 10 Gy.

In-pile test of Li2TiO3 pebble bed with neutron pulse operation

Abstract Lithium titanate (Li 2 TiO 3 ) is one of the candidate materials as tritium breeder in the breeding blanket of fusion reactors, and it is necessary to show the tritium release behavior of Li 2 TiO 3 pebble beds. Therefore, a blanket in-pile mockup was developed and in situ tritium release experiments with the Li 2 TiO 3 pebble bed were carried out in the Japan Materials Testing Reactor. In this study, the relationship between tritium release behavior from Li 2 TiO 3 pebble beds and effects of various parameters were evaluated. The ( R / G ) ratio of tritium release ( R ) and tritium generation ( G ) was saturated when the temperature at the outside edge of the Li 2 TiO 3 pebble bed became 300 °C. The tritium release amount increased cycle by cycle and saturated after about 20 pulse operations.