An overview of plasma facing components (PFCs) in JT-60SA is introduced, including the upper divertor, lower divertor, inboard first wall and outboard first wall. These PFCs are upgraded in four stages during a plasma experiment in JT-60SA. At stage one, only a partial inboard first wall and upper divertor are installed with graphite tiles and inertial cooling. At stage two, the remaining PFCs including the outboard first wall and lower divertor are installed, and the inboard surface is fully covered. At stage three, the vertical targets of the lower divertor are replaced with CFC monoblock targets and water cooling. At stage four, the whole carbon wall is replaced with a tungsten wall, and remote handling for the lower divertor is introduced. The research and development on major issues concerning PFCs are presented. Mock-ups of a CFC monoblock target were fabricated and it showed good heat removal of the expected maximum heat load of 15 MW m−2 without degradation of the performance. Tungsten-coated samples were fabricated on the CFC and they showed no severe damage at surface temperatures of up to 2000 °C. Laser-welded samples showed good appearance and no defects were found by RT.

For a tokamak fusion DEMO reactor with the fusion output of 2.95GW, neutronic and thermal design of blanket is under way to find a feasible blanket concept. For the continuity with the Japanese ITER-TBM options, this study considered water-cooled blanket with solid breeding materials of Li ceramics (Li4SiO4, Li2TiO3 and Li2ZrO3) and Be multipliers (Be and Be12Ti). Based on a neutronics-heat coupled analysis, the tritium breeding ratio was evaluated so as to satisfy constraints of the operating temperature of ≤900°C for Li ceramics and Be12Ti, and ≤600°C for Be. Cooling water condition was assumed to be 23MPa and 290–360°C. The result indicates that surplus tritium production in lower neutron wall load (Pn) blanket compensates a shortfall in higher Pn blanket and thus the overall tritium production can marginally satisfy fuel self-sufficiency.

The first wall (FW) of ITER blanket includes beryllium (Be) armor tiles joined to CuCrZr heat sink with stainless steel cooling tube and backing plate in order to improve plasma performance and reduce thermal stress. Therefore dissimilar materials joints are indispensable for fabricating the high heat flux components. Since these joints must withstand thermal and mechanical loads caused by the plasma and electromagnetic force, it is important to evaluate the strength and thermal fatigue life of dissimilar materials joints. When the dissimilar materials joints are subjected by external force and thermal loading, the stress of the joint may indicate singularity at the interface edge. Since the stress singularity may lower the strength of joints, the singularity is evaluated numerically for the various materials combinations and joint configuration to be used in high heat flux components of fusion reactors in this investigation. Moreover, tensile test and elasto-plastic FEM analysis are performed to investigate the fracture behavior of Be/Cu alloy and stainless steel/ Cu alloy obtained the FW mock-up. The results reveal two singular solutions of type rpj−1 for a half-plane bonded to a quarter-plane joint and the singularity is larger than that of a bonded quarter-planes joint. From the viewpoint of stress singularity, the configuration of bonded quarter-planes joint is better than the half-plane bonded to a quarter-plane joint. The singularity for W/Cu alloy combination is large compared to other combination of materials. Especially the singularity of stainless steel/ Cu alloy is very small. Tensile specimen of Be/CuCrZr joint fractured at the bonding interface due to the stress singularity. For the stainless steel/ Cu alloy, however, the specimens fractured at the Cu alloy region apart from the interface.

Fluid flow and heating tests on the full-scale Test Blanket Module (TBM) First Wall (FW) mock-up under the TBM-relevant operating condition were carried out in an ion beam irradiation system with a pressurized high temperature water loop in JAEA to verify the fabrication process and design of the TBM First Wall and demonstrate its heat removal capability. The TBM FW mock-up is made of reduced activation ferritic/martensitic (RAFM) steel, F82H, and its parts such cooling channels and plates were assembled with Hot Isostatic Pressing (HIP) method. In the fluid flow test, the cooling water at room temperature is supplied to 15 parallel flow paths in the mock-up. The flow distribution in each path inside the First Wall mock-up is compared with a numerical simulation. This result shows that no severe cross-sectional deformation of the entire flow path in the mock-up takes place during the fabrication process. In the heating test, the mock-up was exposed to repetitive hydrogen ion beam irradiation with the maximum heat flux of 0.5MW/m2. No indication of joint defect of the HIP joint like a hot spot was observed during 80 irradiation cycles.

The real-scale component mock-ups have been successfully fabricated for the Test Blanket Module (TBM) with the water cooled ceramic breeder. The TBM has a box structure and contains the breeder and multiplier pebble beds inside. For the side wall and the pebble bed container, fabrication processes have been established and their functions have been confirmed by performance tests. In addition, the side walls and the Ushaped first wall that was previously fabricated have been assembled into an open box with five faces. The achieved technologies can be applied to the fabrication and the assembly of the back wall, and then the box structure of the TBM will promisingly be built in the next stage.

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.

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 bed's 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.

For tokamak fusion DEMO, neutronics and thermal design was carried out to find a blanket concept with reality. For the continuity with the Japanese ITER-TBM options, this study considered water-cooled blanket with solid breeding materials of Li ceramics (Li4SiO4, Li2TiO3 and Li2ZrO3) and Be multipliers (Be and Be12Ti). On the based on multilayered structure with Li4SiO4 pebbles, Be plate and mixture of Li4SiO4/Be12Ti, the local TBR of 1.42 (corresponds to the net TBR of 1.05) was obtained. In addition, it was concluded that in-between conducting shell structure can be placed at rw/a = 1.32-1.35 with satisfying fuel self-sufficiency.

SlimCS is the conceptual design of a compact fusion DEMO plant assuming technologies foreseeable in 2020–2030s. For continuity of blanket technology from the Japanese ITER-TBM, the prime option of blanket is water-cooled solid breeder with Li2TiO3 (or Li4SiO4) and Be. A reduced-activation ferritic–martensitic steel (RAFM) and subcritical water are chosen as the structural material and coolant, respectively. The reactor has somewhat complex torus configuration with a sector-wide conducting plate slipped in between the replaceable (front) and permanent (back) blanket. In order to allow flexibility of maintenance in such a complex configuration, sector transport hot cell maintenance scheme is adopted. This paper describes characteristics of SlimCS with a focus on materials selection.

The first wall (FW) of ITER blanket includes beryllium (Be) armor tiles joined to CuCrZr heat sink with stainless steel (SS) cooling tube and backing plate in order to improve plasma performance and reduce thermal stress. Since these joints must withstand thermal and mechanical loads caused by the plasma and electromagnetic force, it is important to evaluate the strength and thermal fatigue life of the joints. Qualification tests of the FW fabrication technology are performed in order to qualify the joining technologies. The Japan Domestic Agency (JADA) has fabricated the qualification mock-up with a hot isostatic pressing (HIPing) technique and carried out the materials testing and the non-destructive examinations. When the dissimilar materials joints are subjected to external force and thermal load, the stress distribution shows singularity and large stress concentration at the interface edges caused by discontinuity of the elastic properties. The characteristics of stress singularity were analyzed for the Be/CuCrZr and SS/CuCrZr joints. The results of tensile test showed that the Be/CuCrZr joint fractured at the interface because of large stress concentration at the interface edges. For the SS/CuCrZr joint, however, the joint fractured at CuCrZr apart from the interface. There was insignificant stress concentration at the interface edges, because those elastic properties are almost same. Since Charpy impact strength of the SS/CuCrZr joint was inferior to the CuCrZr, elasto-plastic finite element analyses were also carried out to clarify the deformation characteristics and fracture of the tensile and Charpy specimens. Those fracture behavior were caused from the plastic deformation properties of both materials.

In JAEA, the Test Blanket Module (TBM) with water-cooled solid breeder is being developing. This paper presents recent achievements of the research activities for the TBM, focusing on the pebble bed of the tritium breeder materials and tritium behaviour. For the breeder material, the chemical stability of Li2TiO3 has been improved by Li2O additives. In order to analyze the pebble bed behaviour, thermo-mechanical properties of the Li2TiO3 pebble bed has been experimentally obtained. In order to verify nuclear properties of the pebble bed, the activation foil method has been proposed and a preliminary experiment has been conducted. For the tritium behaviour, the chemical densified coating method has been well developed and tritium recovery system has been modified taking account of the design change of the TBM.

Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with an advanced concept with supercritical water in the previous DEMO (DEMO-2001), the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m or lower, which will be a critical constraint for determining a handling power of DEMO (namely, a summation of alpha heating power and external input power for current drive).

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.

This paper presents progresses of the test strategy development, design and supporting R&Ds of solid breeder Test Blanket Modules (TBMs) in Japan. Japan is proposing to test its unique designs of Water Cooled Solid Breeder (WCSB) TBM and Helium Cooled Solid Breeder (HCSB) TBM from the beginning of the ITER operation. Water cooled solid breeder blanket has the better capability of heat removal and applicability to the more compact and economical fusion reactors. On the other hand, if it uses beryllium pebble beds as the neutron multiplier, it has the potential safety concern of thermo-chemical excursion of water and beryllium reaction in the case of the coolant water ingress in the beryllium pebble beds. Safety analysis of the WCSB TBM showed the possible preventive measures to such a case in TBM tests in ITER. Also, to void such potential safety concern, the advanced multiplier material, Be-Ti alloy, is under development in Japan, because it has very low reactivity with water even in high temperature. Helium gas coolant has less heat removal capability but higher safety to thermo-chemical excursion phenomena. As for the design work, structure design showed steady progress and clarified detailed structure taking into account the fabrication procedure. As for supporting R&Ds, the corrosion characteristics of the structural material by high temperature and pressure water was clarified as one of critical structure integrity issues. Also, important design data of the breeder pebble bed has been clarfied. Along with the development progress, the test strategy has been investigated to obtain the most effective results of TBM test program.

Three options of fusion DEMO plant are proposed characterized by functions of the center solenoid (CS). The prime option uses a downsized CS, which does not provide sufficient V-s for plasma current ramp-up but supplies enough coil current for plasma shaping. This option produces a fusion output of 3 GW with a major radius of 5.5 in, aspect ratio of 2.6, normalized beta of 4.3 and maximum field of 16.4 T. The estimated reactor weight is lighter than that of other conventional tokamak reactors, suggesting an economic advantage. The plant uses rather conservative technologies such as Nb3Al superconductor, water-cooled solid breeder blanket, low activation ferritic steel as the structural material and tungsten monoblock divertor plate. The design philosophy and key issues related to the constituent technologies of the plant are described in the present paper. (c) 2005 Elsevier B.V. All rights reserved.

Our idea on the DEMO plant is that it must demonstrate (1) an electric power generation of one GW level, (2) self-sufficiency of tritium fuel (TBR is more than 1.05), (3) year-long continuous operation, etc. At the same time, DEMO is expected to use technologies to be proven by 2020 and present an economical prospect of fusion energy in the operational time of the reactor. The design guidelines for the blanket are defined in order to meet the mission of the DEMO plant as mentioned above. Major design conditions are surface heat flux of 0.5 MW/m(2) with peaking factor of 2, a neutron wall load of 3.5 MW/m(2) with peaking factor of 1.5 and a neutron fluence of about 10MW/m(2).To moderate the technological extrapolation, reduced activation ferritic steel (F82H) structural material, Li2TiO3 and Be neutron multiplier are considered. To improve the economical aspect, supercritical water with inlet/outlet temperatures of 280/510 degrees C is chosen as coolant material, with coolant pressure of 25 MPa. As a result, a thermal efficiency of 41% is achieved. To obtain higher plasma performance, MHD instabilities suppressing shell structure is adopted with structural compatibility to the blanket structure. To meet higher plant availability requirements (more than 75%), the hot cell maintenance approach is selected for the replaceable power core components. (c) 2006 Published by Elsevier B.V.

The concept for a compact DEMO reactor named "SlimCS" is presented. Distinctive features of the concept is low aspect ratio (A = 2.6) and use of a reduced-size center solenoid (CS) which has a function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field (TF) coil system which contributes to reducing the weight and construction cost of the reactor. SlimCS is as compact as advanced commercial reactor designs such as ARIES-RS and produces 1 GWe in spite of moderate requirements for plasma parameters. Merits of low-A, i.e. vertical stability for high elongation and high beta limit are responsible for such reasonable physics requirements. 1. Concept of reactor Conventional tokamak reactor design requires CS with a large bore so that the flux swing capacity ΨCS ( = 2π(RCS-dCS/6)Bpm where RCS, dCS and Bpm are the inner radius and width of CS and the maximum poloidal field, respectively) is prepared for plasma current ramp-up plus extra flux. Incidentally, when tokamak power reactors are operated in the steady state, the most important role of CS is plasma shaping rather than poloidal flux supply. This means that conventional tokamak reactor designs have excess capacity of CS to use just in startup and shutdown of operation. Assuming matured current ramp technology with non-inductive current drive, the CS diameter can be reduced as long as CS produces poloidal magnetic field sufficient for plasma shaping. We have been conceiving the DEMO reactor SlimCS using the reduced-size CS with an outer radius of 0.7 m which has the capability of plasma shaping (triangularity of ~0.4) perhaps enough to obtain high confinement in high density region and possibly to avoid giant edge-localized modes [1]. Such a reduced-size CS allows a small inner leg radius (RTF) of TF coils, eventually contributing to a reduction in the magnetic energy of the TF coil system [2]. As known by the Virial theorem, the weight of a TF coil system increases with its magnetic energy. This is because significant part of the coil system consists of structural material to support the magnetic energy. On top of this, the coil system occupy a central position of the construction cost of reactor. For these reasons, the basic concept of SlimCS, i.e. the reduced-size CS, can eventually have large impact on a reduction in the

A solid breeder blanket was selected as the primary candidate blanket of the fusion power demonstration plant (DEMO plant) in Japan. Test blanket module (TBM) testing in ITER is the most important milestone for the development of the blanket of the DEMO plant. Therefore, the major development effort has been focused on the development of the TBMs in Japan. This paper presents a structural concept of solid breeder TBMs cooled by pressurized water and helium gas. In the sub-module concept which was developed, a slitted box structure was adopted for both TBMs to simulate the DEMO blanket structure and to endure the over-pressurization in case of the coolant ingress to box structure. The breeder and neutron multiplier formed by small pebbles are packed separately in a layered structure. Detailed internal structure was developed with consideration of fabrication procedure, coolant flow route and purge gas flow route. Coolant flow route was considered from the view points of thermal–hydraulic and heat transfer performances. As for He purge gas flow passages, separated passages for tritium breeder layers and neutron multiplier layers were adapted to avoid tritium migration to the neutron multiplier layers.

Since fusion power and neutron fluence of the compact ITER has been reduced, thermal transients due to decay heat of the tokamak components such as in-vessel components and vacuum vessel (VV) were numerically analyzed for the purpose of identifying the necessity of cooling function as the safety measure. The result shows that the maximum VV temperature remains around 500 °C even under the extremely hypothetical conditions, assuming that all the coolants in the VV and in-vessel components are lost instantaneously. In addition, the maximum temperature appears after about 100 days and hence reducing the temperature rise can be practically achieved during such a long grace period. As a whole, it has been clarified that the decay heat removal can be passively achieved by only radiation without any active cooling measures. This paper describes the analysis results on thermal transient due to decay heat, including sensitivity study on the effect of heat connection and removal characteristics on the temperature rise.