M. Akiba

Breeding blanket concepts for fusion and materials requirements

This paper summarizes the design and performances of recent breeding blanket concepts and identifies the key material issues associated with them. An assessment of different classes of concepts is carried out by balancing out the potential performance of the concepts with the risk associated with the required material development. Finally, an example strategy for blanket development is discussed.

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.

Achievements in the development of the Water Cooled Solid Breeder Test Blanket Module of Japan to the milestones for installation in ITER

As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, a water cooled solid breeder (WCSB) TBM is being developed. This paper shows the recent achievements towards the milestones of ITER TBMs prior to the installation, which consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, targeting the detailed design final report in 2012, structure designs of the WCSB TBM and the interfacing components (common frame and backside shielding) that are placed in a test port of ITER and the layout of the cooling system are presented. As for the module qualification, a real-scale first wall mock-up fabricated by using the hot isostatic pressing method by structural material of reduced activation martensitic ferritic steel, F82H, and flow and irradiation test of the mock-up are presented. As for safety milestones, the contents of the preliminary safety report in 2008 consisting of source term identification, failure mode and effect analysis (FMEA) and identification of postulated initiating events (PIEs) and safety analyses are presented.

Design study of National Centralized Tokamak facility for the demonstration of steady state high-β plasma operation

Design studies are shown on the National Centralized Tokamak facility, formerly called JT-60SC. The machine design is carried out to investigate the capability for flexibility in aspect ratio and shape controllability for the demonstration of the high-β steady state operation with nation-wide collaboration, in parallel with ITER towards DEMO. Two designs are proposed and assessed with respect to the physics requirements such as confinement, stability, current drive, divertor and energetic particle confinement. The operation range in the aspect ratio and the plasma shape is widely enhanced consistent with sufficient divertor pumping. Evaluations of the plasma performance towards the determination of the machine design are presented.

Overview of the National Centralized Tokamak programme

An overview is given of the National Centralized Tokamak (NCT) programme as a research programme for advanced tokamak research to succeed JT-60U. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility is pursued in aspect ratio and shape controllability for the demonstration of the high-β steady-state, feedback control of resistive wall modes, wide current and pressure profile control capability and also very long pulse steady-state operation. Existing JT-60 infrastructure such as the heating and current drive system, power supplies and cooling systems will be best utilized for this modification.

Thermal Conductivity Measurement with Silica-Coated Hot Wire for Li4SiO4 Pebble Bed

The effective thermal conductivity of a Li4SiO4 pebble bed was measured by the hot wire method. The bare and silica-coated Nichrome heaters were used as the hot wires. At 975 K, effective thermal conductivity was not measured correctly by the bare hot wire. This is due to the fact that the electrical signal of a bare thermocouple is distorted due to the electrical conductivity of Li4SiO4. Using a silica-coated hot wire, effective thermal conductivity can be measured at temperatures ranging from room temperature to 975 K. The effect of the coating layer on the measured effective thermal conductivity was estimated to be small and corresponded to the experimental data. The hot wire method with silica coating can be applied to other ceramic breeder materials.

Tritium Absorption of CO-Deposited Carbon Films, Graphite and Polycrystalline Tungsten

Abstract Tritium retention in plasma facing materials is a primary issue for ITER and next step fusion devices, since it greatly affects its safety and operational schedule. In the ITER, carbon and tungsten are used as divertor materials. In the present study, co-deposited carbon film, tungsten and isotropic graphite were exposed to tritium gas, and then the amount of absorbed tritium was investigated. During the tritium exposure, the partial pressure of tritium gas was kept at 10 Pa. The sample temperature was kept a constant in the range from RT to 573 K. The amounts of absorbed tritium were evaluated by ²-ray-induced X-ray spectrometry (BIXS). The amounts of absorbed tritium in co-deposited carbon films were one or two orders of magnitude larger than that of polycrystalline tungsten and isotropic graphite. The amount of absorbed tritium for co-deposited carbon film with a high volume density (1.53 g/cm3) was several times larger than that of the film with a low volume density (1.13 g/cm3). The amount of absorbed tritium increased with the temperature. These results indicate that co-deposited carbon films can absorb much larger amount of tritium than tungsten and graphite, and carbon film density affects the amount of absorbed tritium.

Determination of hydrogen diffusion coefficients in F82H by hydrogen depth profiling with a tritium imaging plate technique

Abstract Hydrogen diffusion coefficients in a reduced activation ferritic/martensitic steel (F82H) and an oxide dispersion strengthened F82H (ODS-F82H) have been determined from depth profiles of plasma-loaded hydrogen with a tritium imaging plate technique (TIPT) in the temperature range from 298 K to 523 K. Data of hydrogen diffusion coefficients, D, in F82H are summarized as D [m2 s−1] =1.1×10−7 exp(-16[kJ mol−1]/RT). The present data indicate almost no trapping effect on hydrogen diffusion due to an excess entry of energetic hydrogen by the plasma loading, which results in saturation of the trapping sites at the surface and even in the bulk. In the case of ODS-F82H, data of hydrogen diffusion coefficients are summarized as D [m2 s-1] =2.2×10−7 exp(−30[kJ mol−1]/RT) indicating a remarkable trapping effect on hydrogen diffusion caused by tiny oxide particles in the bulk of F82H.