Koichiro Ezato

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.

Critical Heat Flux in Subcooled Water Flow of One-Side-Heated Screw Tubes

The purpose of the International Thermonuclear Experimental Reactor (ITER) divertor, which is located at the bottom of the vacuum vessel, is to exhaust impurities and their power from the plasma. D...

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.

Recent activities related to the development of the plasma facing components for the ITER and fusion DEMO plant

In the ITER procurement, the Japanese Domestic Agency (JADA) will procure the divertor outer vertical target (OVT). The qualification of the manufacture of the components has been started by validating JADAs technical capability. JADA has developed vertical target qualification prototypes that cover most of the critical technical issues in the series production. The prototypes have been high heat flux tested under effective coordination by the ITER organization (IO) and showed sufficient durability. JADA has successfully obtained the certification to conduct the procurement of the OVT by the IO. Development of a breeding blanket is one of the most important issues to realize the DEMO. Test blanket module (TBM) testing is a key milestone toward the DEMO. In JAEA, R&D on the water-cooled blanket has been performed. As a result, a full-scale TBM first wall mock-up has successfully been developed. This mock-up showed sound thermal performance in preliminary testing.

Assessment of Applicability of Two-Fluid Model Code ACE-3D to Heat Transfer Test of Supercritical Water Flowing in an Annular Channel

A supercritical-water-cooled reactor (SCWR) is a high-temperature, high-pressure water cooled reactor that operates above the critical pressure of water. In order to perform efficiently the thermal design of the SCWR, it is important to assess the thermal hydraulics in rod bundles of the core. Experimental conditions of mockup tests, however, may be limited because of technical and financial reasons. Therefore, it is required to establish an analytical design technique that can extrapolate experimental data to various design conditions of the reactor. Japan Atomic Energy Agency (JAEA) has improved the three-dimensional two-fluid model analysis code ACE-3D, which was originally developed for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water in the supercritical region. In the present study, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which were performed at JAEA, were analyzed with the improved ACE-3D to assess the prediction performance of the code. As a result, it was implied that the ACE-3D code is applicable to the prediction of wall temperatures of a single rod that simulates the fuel bundle geometry of the SCWR core.

ITER Activities in Japan

Japanese contributions to ITER engineering design activities are presented, together with an introduction of the objectives and design of the ITER, whose program has been carried out through international collaboration by the European Union, Japan, Russian Federation, and the United States. New technologies have been produced through the development, fabrication, and testing of scalable models in the fields of superconducting magnets, reactor structures with vacuum vessels, remote-maintenance machines, high-heat-flux plasma facing components, neutral beam injectors, high-power millimetre-wave generators, etc. As major contributions from Japan, development and testing results of a 13-T, 640-MJ, Nb3Sn pulsed magnet; an 18-deg sector of a vacuum vessel with a height of 15 m and a width of 9 m; CFC armor for a CuCrZr cooling tube that withstood 20 MW/m2; a 31 mA/cm2 negative ion beam source; a 1-MeV beam accelerator; and a 1-MW 170-GHz gyrotron are described.

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.

Overview of the R&D Activities of Water Cooled Ceramic Breeder Blanket

This paper overviews the R&D activity of Water Cooled Ceramic Breeder (WCCB) Blanket in Japan. Japan is performing R&D of WCCB Blanket as the primary candidate of the breeding blanket for the fusion DEMO reactor. Regarding the development of blanket module fabrication technology, a real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS) F82H. Using fabricated FW mockup, thermo-hydraulic performance and high heat flux tests were successfully performed with the heat flux equivalent to ITER TBM condition, 0.5 MW/m2 , 80 cycles with the coolant condition as DEMO, 15 MPa 300 °C. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure have been successfully fabricated. Furthermore, assembling of the real scale FW plate mockup and SW plate mockup was successfully performed. Development of major key technologies for the WCCB TBM structure fabrication has been progress toward the establishment of fabrication technology of WCCB TBM.Copyright