Takanori Hirose

Compact DEMO, SlimCS: design progress and issues

The design progress in a compact low aspect ratio (low A) DEMO reactor, SlimCS, and its design issues are reported. The design study focused mainly on the torus configuration including the blanket, divertor, materials and maintenance scheme. For continuity with the Japanese ITER-TBM, the blanket is based on a water-cooled solid breeder blanket. For vertical stability of the elongated plasma and high beta access, the blanket is segmented into replaceable and permanent blankets and a sector-wide conducting shell is arranged inbetween these blankets. A numerical calculation indicates that fuel self-sufficiency can be satisfied when the blanket interior is ideally fabricated. An allowable heat load to the divertor plate should be 8 MW m−2 or lower, which can be a critical constraint for determining a handling power of DEMO.

SlimCS—compact low aspect ratio DEMO reactor with reduced-size central solenoid

The concept for a compact DEMO reactor named SlimCS is presented. Distinctive features of the concept are low aspect ratio (A = 2.6) and use of a reduced-size centre solenoid (CS) which has the function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field coil system which contributes to reducing the weight and perhaps lessening the construction cost. Low-A has merits of vertical stability for high elongation (κ) and high normalized beta (βN), which leads to a high power density with reasonable physics requirements. This is because high κ facilitates high nGW (because of an increase in Ip), which allows efficient use of the capacity of high βN. From an engineering aspect, low-A may ensure ease in designing blanket modules robust to electromagnetic forces acting on disruptions. Thus, a superconducting low-A tokamak reactor such as SlimCS can be a promising DEMO concept with physics and engineering advantages.

Progress of Design and R&D of Water Cooled Solid Breeder Test Blanket Module

Abstract Development of Water Cooled Solid Breeder (WCSB) TBM, the primary candidate of ITER Test Blanket Module (TBM), has been performed in Japan, according to the TBM milestones, which are necessary for acceptance of the TBM in ITER for testing from the first day of plasma operation. The TBM milestones consist of milestones on safety assessment, module qualification and design integration in ITER. For the safety milestones, essential source terms were evaluated, and failure modes and effect analysis (FMEA) was performed. Based on the results of FMEA, safety assessment was performed. For the milestones of the design integration, detailed structural design of the TBM and the interface structure with the ITER test port were performed. Based on the design, performance analysis such as thermo-mechanical analysis in over-pressurization was performed. For the milestones of the qualification of fabrication technology, essential fabrication technology was developed and near full size first wall of the TBM was successfully fabricated and demonstration of the integrity in heat flux equivalent to ITER. The development of the WCSB TBM is showing steady progress toward the installation in ITER.

Effects of tube drawing on structural material for ITER test blanket module

Abstract A detailed study was conducted on the first wall fabrication process using the reduced activation ferritic/martensitic steel that will be used to fabricate ITER test blanket modules. The effects of the tube-drawing process on microstructural and mechanical properties of F82H is one of the most important issues for fabrication of the module. Square tubes with □11 mm x 1.5T (thickness) mm x 3500L (length) mm have been developed by a cold-rolling method. This tube is long enough to fabricate the first wall without any joint in the cooling path. Its surface roughness (Rz) and outer curvature are less than 1 μm and 1.4 mm, respectively. Dimensions were accurate enough to reduce the assembly gap for a Hot Isostatic Pressing (HIP) joint. Although the rolling process introduced an elongated microstructure containing dense precipitates, this anisotropic microstructure was successfully recovered by heat treatments corresponding to that used in the HIP process. This work demonstrated that the drawing process could be applicable to a fabrication process for the breeding-blanket component.

Key achievements in elementary R&D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

This paper presents the significant progress made in the research and development (R&D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li2TiO3 pebble bed under neutron-pulsed operation conditions are summarized.With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150u2009°C followed by normalizing it at 930u2009°C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R&D on the breeder material, Li2TiO3, the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li2TiO3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14u2009MeV neutrons with an accuracy of 10%. The tritium release rate from the Li2TiO3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation.

Mechanical Properties of JLF-1 Reduced-Activation Ferritic Steels

Abstract To improve material properties of reduced-activation ferritic steels, mechanical properties and microstructure of the steels with adjusting minor alloying elements, such as N, B, Ta, and Ti were investigated. If it became necessary to reduce N contents in the steels for nuclear consideration, B-addition would have the potential to produce a steel comparable, at least in terms of mechanical properties before irradiation, to the JLF-1 IEA heat. Increasing the Ta contents could induce further grain refinement in the JLF-1 steel, but had no significant effect on the tensile and impact properties.

Effect of Helium on Irradiation Creep Behavior of B-Doped F82H Irradiated in HFIR

Abstract The diameter of pressurized tubes of F82H and B-doped F82H irradiated up to ~6 dpa have been measured by a non-contacting laser profilometer. The irradiation creep strains of F82H irradiated at 573 and 673K were almost linearly dependent on the effective stress level for stresses below 260 MPa and 170 MPa, respectively. The creep strain of 10BN-F82H was similar to that of F82H IEA at each effective stress level except 294 MPa at 573K irradiation. For 673K irradiation, the creep strain of some 10BN-F82H tubes was larger than that of F82H tubes. However, the generation of ∼300 appm He did not cause a large difference in the irradiation creep behavior at 6 dpa.