A5025591076- Fusion materials and technologies
- Nuclear Materials and Properties
- Nuclear reactor physics and engineering
- Superconducting Materials and Applications
- Magnetic confinement fusion research
- Advanced materials and composites
- High-Temperature Coating Behaviors
- Laser-Plasma Interactions and Diagnostics
- Nuclear Engineering Thermal-Hydraulics
- Metal and Thin Film Mechanics
- Heat transfer and supercritical fluids
- Hydrogen embrittlement and corrosion behaviors in metals
- Heat Transfer Mechanisms
- Heat Transfer and Boiling Studies
- Nuclear and radioactivity studies
- Heat Transfer and Optimization
- Graphite, nuclear technology, radiation studies
- Metal Alloys Wear and Properties
- Nuclear Physics and Applications
- Metallurgical Processes and Thermodynamics
- Subcritical and Supercritical Water Processes
- Particle accelerators and beam dynamics
- Diamond and Carbon-based Materials Research
- Nuclear materials and radiation effects
- Fluid Dynamics and Turbulent Flows
National Institutes for Quantum Science and Technology
2017-2022
National Institute for Fusion Science
2020
Fusion Academy
2018
Japan Atomic Energy Agency
2007-2016
Advanced Science Research Center
1999-2014
Kyushu University
1995-1999
The design progress in a compact low aspect ratio (low A ) DEMO reactor, ‘SlimCS’, and its issues are reported. study focused mainly on the torus configuration including blanket, divertor, materials maintenance scheme. For continuity with Japanese ITER-TBM, blanket is based water-cooled solid breeder blanket. vertical stability of elongated plasma high beta access, segmented into replaceable permanent blankets sector-wide conducting shell arranged inbetween these blankets. numerical...
The current ITER design employs beryllium, carbon fiber reinforced composite and tungsten as plasma facing materials. Since these materials are exposed to high heat fluxes during the operation, it is essential perform flux tests for R&D of components. Static loads corresponding cycling normal estimated be up 20 MW/m2 in divertor targets around 0.5 at first wall ITER. For static testing, electron beam facilities, particle IR heater in-pile have been performed. Another type, more critical...
Power exhaust to the divertor and conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power major radius of 8.5 m, where plasma parameters were revised appropriate impurity seeding scenario. A system code survey Ar suggested volume-averaged density, concentration from main = 205–285 MW. The simulation (SONIC) was performed leg length 1.6 m fixed edge 250 MW total radiation fraction at edge, SOL ( 0.8), as first step investigate size geometry. At...
The concept for a compact DEMO reactor named 'SlimCS' is presented. Distinctive features of the are low aspect ratio (A = 2.6) and use reduced-size centre solenoid (CS) which has function plasma shaping rather than poloidal flux supply. CS enables us to introduce thin toroidal field coil system contributes reducing weight perhaps lessening construction cost. Low-A merits vertical stability high elongation (κ) normalized beta (βN), leads power density with reasonable physics requirements....
At JAEA, a test blanket module (TBM) with water-cooled solid breeder is being developed. This paper presents recent achievements of research activities for the TBM, particularly addressing pebble bed tritium materials and behaviour. For material, chemical stability Li 2 TiO 3 was improved using O additives. To analyse behaviour, thermomechanical properties were assessed experimentally. verify bed's nuclear properties, activation foil method proposed preliminary experiment conducted. reduce...
In order to treat strongly heated, forced gas flows at low Reynolds numbers in vertical circular tubes, the k-ε turbulence model of Abe, Kondoh, and Nagano (1994), developed for turbulent flow between parallel plates with constant property idealization, has been successfully applied. For thermal energy transport, Prandtl number Kays Crawford (1993) was adopted. The capability handle these assessed via calculations conditions experiments by Shehata (1984), ranging from essentially...
As the primary candidate of ITER Test Blanket Module (TBM) to be tested under leadership Japan, a water cooled solid breeder (WCSB) TBM is being developed. This paper shows recent achievements towards milestones TBMs prior installation, which consist design integration in ITER, module qualification and safety assessment. With respect integration, targeting detailed final report 2012, structure designs WCSB interfacing components (common frame backside shielding) that are placed test port...