Mikio Enoeda

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.

Design Improvements and R&D Achievements for Vacuum Vessel and in-Vessel Components Towards ITER Construction

During the preparation of the procurement specifications of ITER for long lead-time items, several detailed vacuum vessel (VV) design improvements are being pursued, such as elimination of the inboard triangular support, adding a separate interspace between inner and outer shells for independent leak detection of field joints, and revising the VV support system to gain more structural performance margin. Improvements to the blanket design are also under investigation, an inter-modular key instead of two prismatic keys and a co-axial inlet?outlet cooling connection instead of two parallel pipes. One of the most important achievements in the VV R&D has been demonstration of the necessary assembly tolerances. Further development of cutting, welding and non-destructive tests for the VV has been continued, and thermal and hydraulic tests have been performed to simulate the VV cooling conditions. In FW/blanket and divertor, full-scale prototypical mock-ups of the FW panel, the blanket shield block, and the divertor components, have been successfully fabricated. These results make us confident in the validity of our design and give us possibilities of alternate fabrication methods.

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.

Effective thermal conductivity of a Li2TiO3 pebble bed for a demo blanket

ABSTRACT In development of the ceramic breeder blanket, the effective thermal conductivity of pebble beds is an important design parameter. For thermo-mechanical design of blanket, pebble beds were investigated used for Li2TiO3 that was a candidate for tritium breeder. Li2TiO3 pebble beds, whose size was 0.28-1.91 mm diameter, were measured on load under no neutron irradiation. The effective thermal conductivity was increased with load increasing was obtained.

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.

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.

Measurement and Numerical Analysis of Welding Residual Stress in Box Structure Employed for Breeder Blanket System of Iter

In order to evaluate the ability to construct a box structure employed for the breeder blanket system of ITER, a real-scale structure of the current Japanese ITER test blanket module was fabricated by electron beam (EB) welding between the first and side walls according to the manufacturing plan, and the residual stresses were measured using X-ray. Also, to examine the reliability of experimental results, thermal elastic-plastic finite element analysis was conducted. At the corner of the box and weld end point, the relatively higher tensile stresses perpendicular to the weld direction were measured. The measured residual stress parallel to the weld line at the mid-point of the weld line had a fair agreement with the computed result. On the other hand, the experimental value perpendicular to the weld line was different from the numerical result, since the residual stress might be released excessively by the grinding of the wide area. Moreover, from the analyses of bead on EB welding for various thickness plates, it was found that the larger compressive stress perpendicular to the weld line would be produced at the surface of thicker plates due to the larger internal mechanical restraint.

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.