Kensuke Mohri

Design Performance of Front Steering-Type Electron Cyclotron Launcher for ITER

Abstract The performance of a front steering (FS)-type electron cyclotron launcher designed for the International Thermonuclear Experimental Reactor (ITER) is evaluated with a thermal, electromagnetic, and nuclear analysis of the components; a mechanical test of a spiral tube for the steering mirror; and a rotational test of bearings. The launcher consists of a front shield and a launcher plug where three movable optic mirrors to steer incident multimegawatt radio-frequency beam power, waveguide components, nuclear shields, and vacuum windows are installed. The windows are located behind a closure plate to isolate the transmission lines from the radioactivated circumstance (vacuum vessel). The waveguide lines of the launcher are doglegged to reduce the direct neutron streaming toward the vacuum windows and other components. The maximum stresses on the critical components such as the steering mirror, its cooling tube, and the front shield are less than their allowable stresses. It was also identified that the stress on the launcher, which yielded from electromagnetic force caused by plasma disruption, was a little larger than the criteria, and a modification of the launcher plug structure was necessary. The nuclear analysis result shows that the neutron shield capability of the launcher satisfies the shield criteria of the ITER. It concludes that the design of the FS launcher is generally suitable for application to the ITER.

Development of blanket box structure fabrication technology

Abstract Fabrication studies have been performed for the first wall and blanket box structure in the fusion experimental reactor designed in Japan. The hot isostatic pressing technique has been proposed as one of the most promising candidate methods for fabricating the first wall. This paper describes the trial fabrication of a half-scale mock-up for part of an outboard module near the midplane, without the internal structure of a breeding region, to investigate its feasibility and to clarify technological issues associated with the proposed fabrication technologies.

Thermal shock fracture of graphite armor plate under the heat load of plasma disruption

Experiments on the thermal shock brittle fracture of graphite plates were performed. Thermal loading which simulated a plasma disruption was produced by an electron beam facility. Pre-cracks produced on the surface propagated to the inside of the specimen even if the thermal stress on the surface was compressive. Two mechanisms are possible to produce tensile stress around the crack tip under thermal shock conditions. Temperature, thermal stress, and the stress intensity factor for the specimen were analyzed based on the finite element method for various heating conditions. The trend of experimental results under the asymmetric heating agrees qualitatively with the analytical results. This phenomenon is important for the design of plasma facing components made of graphite. Establishment of a lifetime prediction procedure including fatigue, fatigue crack growth, and brittle fracture is needed for graphite armors.

Development and design of an ECRF launching system for ITER

The design and development of an EC beam launcher are described for the ITER EC H&CD system for plasma heating and on/off axis current drive. The launcher in an equatorial port (an equatorial EC launcher) has three movable mirrors and can inject the EC wave beam power of 20 MW with continuous toroidal steering of 20–45° (r=0–0.6a). The rf frequency is 170 GHz. The front shield of the equatorial EC launcher is designed, and has three narrow slots for rf injection. A 10 atm resistant torus diamond window for a vacuum and a tritium boundary is developed and introduced to the launcher design. Nuclear analysis by the two-dimensional radiation transport code, DOT3.5 for the equatorial EC launcher was carried out and showed that the present neutron radiation shield structure of the launcher could satisfy the required shielding criteria.

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.

Development of high heat flux component fabrication technology

Diffusion bonding by Hot Isostatic Pressing (HIP) is proposed as one of the candidate methods for assembling the high heat flux components in a fusion reactor, such as the first wall (FW) and divertor/limiter. This paper describes the results of experimental studies performed to confirm the applicability of HIP bonding. Some facesimiles of the FW were made with two HIP bonding techniques. The mechanical properties such as tensile strength, impact value, and low cycle fatigue strength of the bonded part were investigated using diffusion bonded test specimens. The detectability of ultrasonic tests was also studied on them. From the results, a complex FW structure with coolant channels can be easily fabricated by the grooved plate type and rectangular tube type technique, using HIP equipment. The mechanical tests indicate that a joint with a bonding ratio of more than 88% has almost the same tensile properties and fatigue strength as base metal, though about a 20% reduction in impact value occurs in a joint with about a 98% bonding ratio. In ultrasonic testing, defects in joints with less than a 94% bonding ratio could be detected using a normal type probe, and with a focus type probe, defects could be detected in joints with a bonding ratio of less than 90%.

Characteristic evaluation of HIP bonded SS/DSCu joints for surface roughness

Abstract Bonding technologies of SS/DSCu, SS/SS and DSCu/DSCu using HIP method have been developed for the fabrication of the ITER shielding blanket first wall. In this study, the effect of the roughness of the HIP bonding surface has been studied for the 1–40 μm roughness from the view points of bondability of surfaces by HIPping and simplification of fabrication procedure. For the tensile properties, it has been found that the effect of the roughness within above range is negligibly small. On the other hand, it has been confirmed that the finer the surface roughness, the higher impact properties for the DSCu/DSCu joints. Significant differences are not found in the SS316L/DSCu and SS316L/SS316L joints with 1–40 μm, though the finer the surface roughness, the slightly higher the impact energy value. Also, it has been found that the HIP bonded joint with good bondability is obtained by using higher HIP pressure, i.e. 200 MPa even with coarser surface roughness, i.e. ∼10 μm, to reduce the machining requirement in the fabrication procedure. In the fabrication of the International Thermo-nuclear Experimental Reactor (ITER) shielding blanket, conditions of the surface roughness and the Hot Isostatic Pressing (HIP) pressure will be selected by taking into account the machining cost for surface preparation and the HIP process cost based on dimensions of the part to be HIP bonded.

DEVELOPMENT OF FABRICATION TECHNOLOGIES FOR PLASMA FACING COMPONENTS OF ITER

In the ITER divertor design, a bonding structure of carbon composites (CFCs) and metal materials has been considered as the divertor plate. Thermal stress after brazing due to the mismatch of the thermal expansions is one of the most important problems for the divertor fabrication. As the results of the experimental and numerical studies on several bonding structures, it turned out a monoblock type armor, which has no opening edges at the bonding surface, was preferable to reduce the residual stress. After the optimization of brazing conditions, small divertor mockups have been successfully fabricated with various kinds of CFCs. The support structure development is also one of the major concerns of the divertor plate. A sliding rail structure was proposed to reduce thermal stresses and deformations of the divertor plate due to the thermal expansion. After being performed two- and three-dimensional thermal stress analyses and production trials, a divertor mockup of 1m long has been fabricated with the sliding rail structures. For the first wall fabrication, the manufacturing technique of a corner part is complicated and important. Bending tests of cooling tubes and substrates, and production trials of small specimens have been performed. Based on these results, a first wall mockup with a corner has been successfully fabricated using a HIP bonding technique.

DEVELOPMENT OF FIRST WALL WITH RADIATIVE COOLED GRAPHITE ARMOR TILE

First walls in the next step tokamak reactors such as ITER and FER will be covered with carbon based armors. The first wall with radiative cooled graphite armor tiles is one of the primary candidate concepts in the reactors. Design works and developments of fabrication techniques for this type of first wall were performed. A real scale partial mock-up was fabricated, and provided as a test specimen for heat flux tests performed in the electron beam test stand in JAERI(JEBIS). Analytical studies to expect the characteristics of the mock-up in the heat flux tests were also carried out.