Kunihiko Tsuchiya

Fabrication development and preliminary characterization of Li2TiO3 pebbles by wet process

Lithium titanate (Li 2 TiO 3 ) has attracted attention of many researchers because of easy tritium recovery at low temperature, high chemical stability, etc. The application of small Li 2 TiO 3 spheres has been proposed in some designs of fusion blanket. Although, the wet process and sol-gel method are the most advantageous as a fabrication method of Li 2 TiO 3 pebbles from points of mass production, and of reprocessing necessary for effective use of resources and reduction of radioactive wastes. However, the fabrication of Li 2 TiO 3 pebbles by the wet process has not been established. Therefore, in this study, fabrication development and preliminary characterization of Li 2 TiO 3 pebbles by the wet process were performed, noting the aging and sintering conditions in the fabrication process of gel-spheres. At the best condition. Li 2 TiO 3 pebbles with the target density of 80-85%T.D. were obtained.

Density dependence on thermal properties of Li2TiO3 pellets

Lithium ceramics have been considered as tritium breeders for fusion reactors. As a part of the study program for fusion blanket development, the characteristics of tritium breeders are estimated for the material selection and fusion blanket design. Among the candidate tritium breeders, lithium titanate (Li 2 TiO 3 ) is regarded as one of the best materials because of many advantages, reasonable lithium atom density, low activation, excellent tritium release characteristic at low temperature, compatibility with structural materials, reprocessing, etc. However, only limited data on the thermal and mechanical characteristics of Li 2 TiO 3 have been obtained for breeder material selection and fusion blanket design. In this study, Li 2 TiO 3 pellets with three different densities, i.e., 73%T.D., 83%T.D. and 93%T.D., were prepared. Then, the density dependence and the thermal hysteresis of their thermal conductivity, specific heat and thermal expansion were investigated.

Mechanical properties of CuCrZr alloy and SS316 joints fabricated by friction welding method

Copyright (c) 1996 Elsevier Science B.V. All rights reserved. Copper alloys with high-strength and high-conductivity are being considered for several magnetic fusion energy applications such as the first wall in high power-density devices, resistive magnetic coils, and high-heat flux components. For example, the stainless steel is a structural material while Cu-alloy acts as a heat sink material for the surface heat flux in the first wall. Therefore, development of reliable joints between Cu-alloys and stainless steel (SS316r is required. In the present work, joining tests on Cu−1%Cr−1%Zr/SS316 by friction welding were performed, and optimum fabricating conditions of the Cu-alloy/SS316 joint were determined. Additionally, the characteristics of tensile strength, hardness, metallographical observation and SEM/EPMA analyses on Cu−1%Cr−1%Zr/SS316 fabricated by friction welding were evaluated.

Conceptual Design of Superconducting Magnet System for JT-60SA

The upgrade of JT-60U magnet system to superconducting coils (JT-60SA) has been decided by both parties of Japanese government (JA) and European commission (EU) in the framework of the Broader Approach (BA) agreement. The magnet system for JT-60SA consists of 18 toroidal field (TF) coils, a Central Solenoid (CS) with four modules, seven Equilibrium Field (EF) coils. The TF case encloses the winding pack and is the main structural component of the magnet system. The CS consists of independent winding pack modules, which is hung from the top of the TF coils through its pre-load structure. The seven EF coils are attached to the TF coil cases through supports which include flexible plates allowing radial displacements. The CS modules operate at high field and use Nb3 Sn type superconductor. The TF coils and EF coils use NbTi superconductor. The magnet system has a large heat load from nuclear heating from DD fusion and large AC loss. This paper describes the technical requirements, the operational interface and the outline of conceptual design of the superconducting magnet system for JT-60SA.

The Manufacturing of the Superconducting Magnet System for the JT-60SA

JT-60SA is the satellite tokamak for ITER in the Broader Approach agreement. The JT-60SA uses 18 toroidal field coils, a central solenoid with 4 modules, and 6 equilibrium field coils, they are all superconducting coils with forced flow cooled conductors. All detailed designs of these superconducting coils have been completed. The manufacturing of conductors and coils are progressing in Japan and EU. This paper shows the latest manufacturing activities and final design adjusting of its magnet system and their utilities.

Conductor Design of CS and EF Coils for JT-60SA

The conductor for central solenoid (CS) and equilibrium field (EF) coils of JT-60 Super Advanced (JT-60SA) were designed. The conductor for CS is Nb3Sn Cable-In-Conduit (CIC) conductor with JK2LB jacket. EF coil conductors are NbTi CIC conductor with SS316LN jacket. The field change rate (3.9 T/s), faster than ITER generates the large AC loss in conductor. The analyses of current sharing temperature (Tcs)margins for these coils were performed by the one-dimensional fluid analysis code with transient heat loads. The margins of these coils are 1 K for the plasma standard and disruption scenarios. The minimum Tcs margin of CS conductor is 1.2 K at plasma break down (BD). The margin is increased by decreasing the rate of initial magnetization. It is found that the disruption mainly impacts the outer low field EF coil. The disruption decreases the Tcs margin of the coil by >1 K. A coupling time constant of <100 ms, Ni plating, and a central spiral are required for NbTi conductor.

Development of advanced tritium breeders and neutron multipliers for DEMO solid breeder blankets

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.

Lithium reprocessing technology for ceramic breeders

Abstract Lithium ceramics have been receiving considerable attention as tritium breeding materials for fusion reactors. Reprocessing technology development for these materials is proposed to recover lithium, as an effective use of resources and to remove radioactive isotopes. Four potential ceramic breeders (Li 2 O, LiAlO 2 , Li 2 ZrO 3 and Li 4 SiO 4 ) were prepared in order to estimate their dissolution properties in water and various acids (HCl, HNO 3 , H 2 SO 4 , HF and aqua regia). The dissolution rates were determined by comparing the weight of the residue with that of the starting powder (the weight method). Recovery properties of lithium were examined by the precipitation method.

Development and Test of JT-60SA Central Solenoid Model Coil

A central solenoid (CS) model coil (CSMC) was manufactured by using real manufacturing jigs and procedure to validate the CS manufacturing processes for JT-60SA. The winding accuracy and the temperature control precision during the heat treatment met the requirements. The vacuum pressure impregnation process was also successfully finished. The cold test of the CSMC was performed as a final check of the manufacturing process. The joint resistance, the Ic, and the pressure drop measurements were conducted as the verification test. The results of verification test satisfied the design requirements. These results indicate that the manufacturing processes of the JT-60SA CS has been established. The manufacturing of real CS pancakes just started after finishing the CSMC test.

R&D of a Li2TiO3 pebble bed for a test blanket module in JAEA

At JAEA, a test blanket module (TBM) with a water-cooled solid breeder is being developed. This paper presents recent achievements of research activities for the TBM, particularly addressing the pebble bed of the tritium breeder materials and tritium behaviour. For the breeder material, the chemical stability of Li2TiO3 was improved using Li2O additives. To analyse the pebble bed behaviour, thermomechanical properties of the Li2TiO3 pebble bed were assessed experimentally. To verify the pebble beds nuclear properties, the activation foil method was proposed and a preliminary experiment was conducted. To reduce the tritium permeation, the chemical densified coating method was developed and the coating was attached to F82H steel. For tritium behaviour, the tritium recovery system was modified in consideration of the design change of the TBM.