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The current status and the progress of R&D activities for a Fusion DEMO reactor in the QST Rokkasho Fusion Institute is reported. In order to advance the JA DEMO activity, not only Japanese domestic activity but also international collaborations of Broader Approach Activity and ITER related activities are conducted in the QST Rokkasho Fusion Institute. Activities for DEMO design and relevant R&D, design of a fusion neutron source and development of an accelerator, ITER-TBS, tritium handling technology and information technology infrastructures including a supercomputer system and a remote experimentation system are carried out for a Fusion DEMO reactor.

The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokka...

JA DEMO selected the vertical maintenance scheme as the primary maintenance option. In order to improve reliability of Remote Handling (RH) and plant availability, update of in-vessel transferring mechanism of the segment, pipe unit structure on maintenance port were carried out. The RH equipment is composed of end-effectors (grippers) for the banana segment, a power manipulator, a telescopic guide and a carrier. The pipe structure in vertical maintenance port was updated considering the thermal expansion, easy handling (short maintenance time) and alignment of welding groove. Based on estimation of maintenance time, four works in parallel will be required for plant availability reachable for commercialization (>60%).

JA DEMO selected the vertical maintenance scheme as the primary maintenance option. In order to improve reliability of Remote Handling (RH) and plant availability, update of in-vessel transferring mechanism of the segment, pipe unit structure on maintenance port were carried out. The RH equipment is composed of end-effectors (grippers) for the banana segment, a power manipulator, a telescopic guide and a carrier. The pipe structure in vertical maintenance port was updated considering the thermal expansion, easy handling (short maintenance time) and alignment of welding groove. Based on estimation of maintenance time, four works in parallel will be required for plant availability reachable for commercialization (>60%).

Power exhaust scenario and divertor design for a steady-state Japan (JA) DEMO and a pulse Europe (EU) DEMO1 have been investigated as one of the most important common issues in Broader Approach DEMO Design Activity. Radiative cooling is a common approach for the power exhaust scenario. For the JA DEMO, development of the divertor design appropriate for high Psep/Rp ~30 MWm-1 is required, while radiation fraction in the main plasma (fradmain = Pradmain/Pheat) is ITER-level (0.40-0.45) and the exhaust power above the L to H-mode power threshold (fLH = Psep/PthLH) is large margin (~2). For the EU DEMO1, larger fradmain (= 0.67) and smaller fLH (=1.2) plasma is required, using higher-Z impurity seeding, in order to apply ITER-level devertor (Psep/Rp = 17 MWm-1). ITER technology, i.e. water cooling with W-monoblock and Cu-alloy (CuCrZr) heat sink, is a baseline for JA and EU to handle the peak heat load of 10 MWm-2-level, and neutron flux and dose are comparable. For the JA DEMO, two different water-cooling pipes, i.e. CuCrZr and F82H steel, are proposed. For the EU DEMO1, heat sink consists of all Cu-alloy pipe, and several choices of the heat sink components have been developed appropriate to the high neutron irradiation condition. These JA and EU approaches of the power exhaust scenario will provide important case-studies for the future decision of the DEMO divertor design.

The requirements for Japanese DEMO are to demonstrate (1) steady and stable electrical power generation, (2) plant availability on a level relevant for commercialization, and (3) self-sufficiency of fuels. The DEMO design activity has been mainly focused on challenging DEMO design issues such as divertor heat removal, superconducting magnets, breeding blankets, remote maintenance and the water cooling system, in order to reach the systemic consistency toward a promising DEMO concept. The current status of the DEMO conceptual design activity is presented.

The divertor plasma performance and the power handling are studied for 8 m class DEMO reactor with the fusion power of 1.5 GW. Due to the high impurity radiation power (80% of the exhausted power), the full detachment at the inner target and the partial detachment at the outer target are obtained for a relatively low electron density of 1.8 × 1019 m−3 at the outer mid-plane separatrix. The SONIC simulation shows the target heat load less than 8 MW/m2, which can be handled by the ITER-like divertor target, for both target. However, at the outer target, the ion temperature is still high which may cause the target erosion. For the divertor power handling and suppression of the target erosion, the divertor design study have to be further proceeded as well as the core plasma design. Dependence of the mid-plane separatrix density and the impurity concentration on the fuel gas puff is also studied. With increasing fuel gas puff rate, the mid-plane separatrix density increases and the Ar impurity concentration at the outer mid-plane decreases to 0.5%.

Requirements for pellet injection system for a Japanese DEMO concept are comprehensively assessed in order to elucidate direction and target of R&D. A fusion output control simulation study using 1.5D transport code is carried out to estimate requirement of fuel particle source. Next, multivariable parameter dependence survey of the fuel deposition source on pellet speed, poloidal injection angle, pellet mass and pedestal height is performed. The conditions to achieve the target fueling depth derived from the output control study are shown. The AUG calibrated relation is referred to consider spatial restriction of tokamak geometry which relates maximum pellet velocity to curvature of pellet guide tube. Then the restriction is collated to the results of multivariable survey to find the most advantageous pellet speed, injection angle, and pellet mass which is consistent with both output control and tokamak geometry. Consequently, a conceptual design of fuel pellet injection system as the R&D target toward to DEMO is proposed.

The Remote Experimentation Centre (REC) in Japan has been preparing to replicate the full dataset of ITER over 10 000 km distance. In such a multi-site data repository environment, the data location informing service will be essential to find and retrieve the data efficiently. Considering the long latency time and the self sustainability of remote sites, the data location database should be running at each repository site. Multi-master asynchronous replication between cooperating databases will be essential to realize the remote experimental collaborations in fusion research. This study has investigated the functional differences of some relational databases and found that Postgres BDR has the expected database replication capabilities. Bi-directional replication (BDR) tests by using the LHD database and SNET revealed that the throughputs are sufficient for remote collaborations in fusion experiments.

Technical verification has been progressing for high efficiency data replication between ITER and the Remote Experimentation Centre (REC) in Japan. Transferring a huge amount of data simultaneously to multiple destinations might cause excessive loads and network bandwidth on the sender so that daisy-chained relay transfer would be a considerable solution. This study demonstrates how efficiently the replication relay could be realized for the next-generation fusion experiments, such as ITER and JT-60SA. All the LHD data were consecutively sent to the REC through the global loop path (GLP: Toki - Gifu - Tokyo - Amsterdam - New York - Los Angeles - Tokyo - Aomori - Rokkasho) on SINET5 L2VPN, whose round-trip time is almost 400 ms. MMCFTP was used for the data transferring application. In both the Japan domestic path and the GLP cases, every transfer shows a very stable flattop speed as the preset 8 Gbps. However, longer gap times were needed in MMCFTP initial negotiation to establish numerous sessions. The performance optimized NVMe and iSCSI striped storages have shown higher throughputs than the ITER estimated initial data rate of 2 GB/s. Those knowledge enable the design optimization of not only the sender/receiver servers with their storages but also the intermediate relay server system.

In the DEMO divertor, the neutral density becomes high to produce the full detachment and therefore the photon trapping can become important. In this paper, effects of the photon trapping on the DEMO divertor plasma has been studied. The pre-evaluation of the photon trapping effects on the fixed background plasma profile was carried out by using an iterative self-consistent collisional radiative model. The recombining plasma near the inner target and the private region changed to the ionizing plasma by the photon-excitation. Based on the preevaluation result, the database of the effective ionization rate coefficient including the photon transport inside a 2 mm sphere. Advantage of the 2 mm sphere approximation is that the extra calculation cost is not necessary. Iterative calculation of the SONIC including the photon trapping effects was carried out. While the electron density increased and the neutral density decreased in the wide region, the electron density decreases close to the inner strike point. This may be due to decrease in the ionization rate by decrease in the neutral density. (© 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

In order to realize a feasible DEMO, we designed an in-vessel component including the conducting shell. The project is affiliated with the broader approach DEMO design activities and is conceptualized from a plasma vertical stability and engineering viewpoint. The dependence of the plasma vertical stability on the conducing shell parameters and the electromagnetic force at plasma disruption were investigated in numerical simulations (programmed in the 3D eddy current analysis code and a plasma position control code). The simulations assumed the actual shape and position of the vacuum vessel and in-vessel components. The plasma vertical stability was most effectively maintained by the double-loop shell.

The blanket modules in DEMO are made of reduced-activation ferritic martensitic (RAFM) steel F82H. This material is ferromagnetic and it drags the magnetic field lines into the first wall (FW). Because of this, the heat load by the plasma heat flux, which goes along the magnetic field line will become higher. In this research, the first analysis of such effect has been done. The extra magnetic field Bm made by RAFM wall becomes higher at inner midplane, and the heat load at the module front surface becomes 1.3 MW/m2 to 5 MW/m2. Additionally, near the toroidal gaps, BM becomes high. Thus, at the top of the FW, magnetic field lines are dragged into the toroidal gaps directly because, the magnetic flux surface is not closed. This makes high (about 10MW/m2) heat load concentration at the moduel edge. The effect of the NBI port is also analyzed. Also near the port, Bm becomes high and the orbit of the magnetic field lines are changed. The effect of this doesn't occur near the port, but far region such as inner midplane or top of the FW. The heat load becomes 6 MW/m2 at inner midplane. These results indicate that the effect of RAFM steel on the FW heat load is not negligible, and more detailed analysis is necessary.

The blanket modules in DEMO are made of reduced-activation ferritic martensitic (RAFM) steel F82H. This material is ferromagnetic and it drags the magnetic field lines into the first wall (FW). Because of this, the heat load by the plasma heat flux, which goes along the magnetic field line will become higher. In this research, the first analysis of such effect has been done. The extra magnetic field Bm made by RAFM wall becomes higher at inner midplane, and the heat load at the module front surface becomes 1.3 MW/m2 to 5 MW/m2. Additionally, near the toroidal gaps, BM becomes high. Thus, at the top of the FW, magnetic field lines are dragged into the toroidal gaps directly because, the magnetic flux surface is not closed. This makes high (about 10MW/m2) heat load concentration at the moduel edge. The effect of the NBI port is also analyzed. Also near the port, Bm becomes high and the orbit of the magnetic field lines are changed. The effect of this doesn't occur near the port, but far region such as inner midplane or top of the FW. The heat load becomes 6 MW/m2 at inner midplane. These results indicate that the effect of RAFM steel on the FW heat load is not negligible, and more detailed analysis is necessary.

Remote experiment with WEST tokamak in CEA Cadarache from ITER Remote Experimentation Centre (REC) in Rokkasho had been successfully carried out under the agreement between the Government of Japan and the European Atomic Energy Community for the joint implementation of the Broader Approach (BA) activities in the field of fusion energy research. The WEST and the REC sites were connected with the international broadband network for science and education of SINET, GEANT and RENATER. Due to a long distance over 10,000 km of two sites, the latency was about 200 ms (round trip time) in total via the direct line between EU and Japan of SINET5. Based on the communication between local and remote participants, plasma discharges were successfully executed repeatedly according to the pulse schedule produced by remote Session Leader after the approval by Engineer in Charge on WEST site. Intershot analyses were successfully executed during the pulse interval on an analysis server prepared in Partners network in the WEST for external collaborator. Full function required for the remote experiments were successfully demonstrated in the large size tokamak facility, WEST. For remote participation of ITER, however, it is expected that the amount of experimental data and the number of remote participations will become very large. Distributed database system based on data replication using high-speed data transferring method and distributed computing resources in local and remote sites are proposed as a possible solution in ITER.

The conceptual design of Japan’s fusion demonstration plant (JA DEMO) is now being developed. In this paper, an overall plant system concept related to tritium handling in the water-cooling system is developed to give a concrete shape to the present JA DEMO concept as an electric power plant. The basic condition of tritium permeation from the in-vessel components to the primary cooling system is evaluated to be 5.7 g-T/day. The tritium concentration of the primary coolant is assumed to be 1 TBq/kg similar to the heavy water reactor condition. The capacity of the water detritiation system (WDS) is assessed, and the bypass feed water from the primary cooling loop is evaluated to be 94 kg/h under the tritium extraction efficiency of 0.96. Based on those specific parameters, the existing WDS in the heavy water reactor is found to be applicable to that of JA DEMO. Configuration of the primary heat transfer system (PHTS) is also discussed. Based on the heavy water reactor experience, tritium permeation through a steam generator (SG) to the secondary cooling system in PHTS is evaluated at 11.77TBq/year/loop (318 Ci/year/loop), which is found to be less than the restricted amount of tritium disposal for a pressurized water reactor in Japan. The key effect of the heavy water reactor experience is reduction of tritium permeation by oxide layer formed on SG pipes. Finally, a confinement concept of tritium release from PHTS is discussed under the condition of ex-vessel loss of coolant accident (LOCA). A pressure suppression system is installed to prevent the upper tokamak hall from pressurizing at the ex-vessel LOCA, and the tritium leakage from the upper tokamak hall is consequently restrained. The resultant early public dose at the plant site boundary can be reduced to 1.8 mSv, which is negligibly smaller than 100 mSv of the no-evacuation limit recommended by IAEA.

Preceding study (Homma Y, Hoshino K, Tokunaga S, et al., 2018 Contrib. Plasma Phys. 58, 629-637) presented an advanced model for thermal force whose applicable condition is extended from collisional to relatively low collisional plasma. According to this extended model, the thermal force reduces when plasma collisionality decreases. In the present study, the extended thermal force model has been implemented into a SOL-divertor integrated simulation code SONIC, in order to study impact on impurity transport in DEMO-relevant Scrape-off layer (SOL) plasma, due to thermal force reduction. A set of test simulation has been carried out supposing a reference steady- state operation scenario of Japanese DEMO fusion reactor concept (JA DEMO). The thermal force has reduced by as much as 20-70%, reflecting relatively lower collisionality in DEMO SOL plasma. The simulation results have demonstrated that introduction of the collisionality dependence of thermal force leads to as much as 20-80% of effective decrease in impurity density and its content ratio widely over the SOL upstream area under DEMO relevant condition. Compared to the case with conventional thermal force model, relative change rate of impurity content at representative poloidal positions are as follows: low-eld side(LFS) X-point -59%; poloidal top area -22%; HFS upstream area up to -80%; no major impact around LFS upstream.

Plasma heat flux in the peripheral plasma reaches the first wall (FW) along a magnetic field line, and it sometimes causes several MW/m2 orders of magnitude high heat flux concentration at narrow region, such as the edge of the blanket module. Thus, to assess and to reduce the heat load is key issue in the DEMO design activity. In this research, a new heat load analysis code is introduced based on the e-folding model. In this code, the decay length λ is changed depending on wall connection length of magnetic field line (defined as the length of magnetic field line from the wall to the wall), and parallel heat flux q / / is calculated in each flux tube. This code can calculate the FW heat load simulating actual blanket module shapes. The 0.23 MW/m2 peak heat load at inner midplane in the case of the ideal FW (without gaps between blanket modules) is increased to 22 MW/m2 at toroidal module edge in the case of box shaped module. To shadow the edge and reduce such peak heat load, toroidal, and poloidal roof shaping is applied. Required roof height is analyzed from this code calculation. After shaping, peak heat load is reduced to 1.2 MW/m2. This value is under the allowable value 1.5 MW/m2, and in this case, surface temperature is also less than allowable temperature.

Plasma heat flux in the peripheral plasma reaches the first wall (FW) along a magnetic field line, and it sometimes causes several MW/m2 orders of magnitude high heat flux concentration at narrow region, such as the edge of the blanket module. Thus, to assess and to reduce the heat load is key issue in the DEMO design activity. In this research, a new heat load analysis code is introduced based on the e-folding model. In this code, the decay length λ is changed depending on wall connection length of magnetic field line (defined as the length of magnetic field line from the wall to the wall), and parallel heat flux q// is calculated in each flux tube. This code can calculate the FW heat load simulating actual blanket module shapes. The 0.23 MW/m2 peak heat load at inner midplane in the case of the ideal FW (without gaps between blanket modules) is increased to 22 MW/m2 at toroidal module edge in the case of box shaped module. To shadow the edge and reduce such peak heat load, toroidal, and poloidal roof shaping is applied. Required roof height is analyzed from this code calculation. After shaping, peak heat load is reduced to 1.2 MW/m2. This value is under the allowable value 1.5 MW/m2, and in this case, surface temperature is also less than allowable temperature.

Remote experiment with WEST tokamak in CEA Cadarache from ITER Remote Experimentation Centre (REC) inRokkasho had been successfully carried out under the agreement between the Government of Japan and theEuropean Atomic Energy Community for the joint implementation of the Broader Approach (BA) activities in thefield of fusion energy research. The WEST and the REC sites were connected with the international broadbandnetwork for science and education of SINET, GÉANT and RENATER. Due to a long distance over 10,000 km oftwo sites, the latency was about 200 ms (round trip time) in total via the direct line between EU and Japan ofSINET5. Based on the communication between local and remote participants, plasma discharges were success-fully executed repeatedly according to the pulse schedule produced by remote Session Leader after the approvalby Engineer in Charge on WEST site. Intershot analyses were successfully executed during the pulse interval onan analysis server prepared in Partners network in the WEST for external collaborator. Full function required forthe remote experiments were successfully demonstrated in the large size tokamak facility, WEST. For remoteparticipation of ITER, however, it is expected that the amount of experimental data and the number of remoteparticipations will become very large. Distributed database system based on data replication using high-speeddata transferring method and distributed computing resources in local and remote sites are proposed as apossible solution in ITER.

We have considered a strategy for reducing the radioactive waste generated by the replacement of in-vessel components, such as blanket segments and divertor cassettes, for the fusion DEMO reactor. ...

A short super-X divertor (SXD) is proposed as an option for the Demo divertor, where the field line length from the divertor null to the outer target was largely increased compared to a similar-size conventional divertor. Physics and engineering design studies for a 3 GW-level fusion power Demo reactor (SlimCS) (Tobita et al., 2009) have recently progressed. Minimal number of the divertor coils were installed inside the toroidal field coil, i.e. interlink-winding. Arrangement of the poloidal field coils and their currents were determined, taking into account of the engineering design such as vacuum vessel and the neutron shield structures, and the divertor maintenance scenario. Divertor plasma simulation showed that significant radiation region is produced between the super-X null and the target. Radiation loss in the divertor was increased, producing fully detached plasmas efficiently. Advantages of the short SXD were demonstrated, but the total peak heat load was a marginal level (10 MW m −2 ) for the engineering design.

Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor.

Vertical instability of an elongated plasma and its stabilized concepts by in-vessel components and vacuum vessel (VV) design have been studied intensely in JA and EU Broader Approach, DEMO Design Activity. The vertical stabilization of the plasma represents one of the key issues for EU and JA DEMO, due to the large distance of the active control coils for the presence of thick breeding blanket system. A feasible DEMO reactor that maintains plasma vertical stability was proposed from an engineering viewpoint. The vertical stability performances are acceptable, without considering a passive stabilizer, if a maximum elongation of κ95 = 1.6 is chosen. For the higher-elongated plasmas (κ95 > 1.70), additional inboard passive stabilizer is effective.

Relations between a net electrical output power and dimensions of components in radial build are investigated based on the ITER plasma performance to develop a conceptual design of DEMO with the net electrical output power of several hundred MW. Reducing the dimensions of in-vessel components and increasing the thickness of the toroidal field coil contribute to strengthen the toroidal magnetic field at plasma, which brings about increase in a net electrical output power. The relation between the minimum plasma major radius and the maximum net electrical output power is clarified. Furthermore effects of improvements in the ITER plasma performance on the net electricity are also analyzed; indicating the increase of normalized beta could have advantage from the viewpoint of the divertor heat load because the increase of synchrotron radiation loss power contributes to reduce the divertor heat load, though the higher energy confinement is required.

Goals of Japan’s fusion demonstration (DEMO) reactor are to demonstrate (1) steady and stable electric power generation in a power plant scale, (2) self-sufficient production of fuel (tritium), and (3) reasonable availability using a remote maintenance scheme anticipated in a commercial plant. Main design parameters of JA DEMO are a plasma major radius of 8.5 m, fusion output of 1.5-2 GW, magnetic field on the plasma axis of 5.94 T. The superconducting coil system of the reactor consists of a central solenoid (CS), 7 poloidal field (PF) coils and 16 toroidal field (TF) coils. Regarding CS and PF coils, superconducting coil technology on DEMO is basically the same as that on the world largest fusion experimental reactor called ITER. In contrast, TF coils have a technology gap on magnetic energy and the resulting stress between ITER and DEMO due to their size and magnetic field. In particular, the necessity of higher design stress is critical for TF coils, requiring the development of high strength cryogenic steels surpassing the existing ones. The fundamental design strategy to mitigate tolerances in TF coil fabrication is also presented.

Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of = 205–285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of = 250 MW and the total radiation fraction at the edge, SOL and divertor ( = 0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load () at the attached region was reduced to ~5 MW m−2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak was less than 10 MW m−2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5–1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak of 10 MWm−2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m−2 was distributed in the major part of the coolant pipe. These results were acceptable for the plasma facing and structural materials.

Power exhaust for a 3?GW class fusion reactor with an ITER-sized plasma was investigated by enhancing the radiation loss from seeding impurity. The impurity transport and plasma detachment were simulated under the Demo divertor condition using an integrated divertor code SONIC, in which the impurity Monte-Carlo code, IMPMC, can handle most kinetic effects on the impurity ions in the original formula. The simulation results of impurity species from low Z (neon) to high Z (krypton) and divertor length with a plasma exhausted power of 500?MW and radiation loss of 460?MW, and a fixed core?edge boundary of 7???1019?m?3 were investigated at the first stage for the Demo divertor operation scenario and the geometry design. Results for the different seeding impurities showed that the total heat load, including the plasma transport and radiation , was reduced from 15?16?MW?m?2 (Ne and Ar) to 11?MW?m?2 for the higher Z (Kr), and extended over a wide area accompanied by increasing impurity recycling. The geometry effect of the long-leg divertor showed that full detachment was obtained, and the peak qtarget value was decreased to 12?MW?m?2, where neutral heat load became comparable to and due to smaller flux expansion. Fuel dilution was reduced but was still at a high level. These results showed that a divertor design with a long leg with higher Z seeding such as Ar and Kr is not fulfilled, but will be appropriate to obtain the divertor scenario for the Demo divertor. Finally, influences of ? and D? enhancement were seen significantly in the divertor, i.e. the radiation and density profiles became wider, leading to full detachment. Both qtarget near the separatrix and Te at the outer flux surfaces were decreased to a level for the conventional technology design. On the other hand, the problem of fuel dilution became worse. Extrapolation of the plasma transport coefficients to ITER and Demo, where density and temperature will be higher than ITER and edge-localized modes are mitigated, is a key issue for the divertor design.

The mechanism of internal transport barrier (ITB) collapse in the reversed magnetic shear configuration is investigated using a global ion temperature gradient (ITG) driven drift wave turbulence code. A heating source and a toroidal momentum source are introduced to follow the self-consistent evolution of the ion temperature and flow profiles. A scenario of transport barrier collapse driven by a meso-scale mode excited in the barrier region is suggested. The importance of the quasi-linear effect due to profile modification as well as three-wave coupling is clearly shown by means of energy transfer analysis. The effect of the toroidal flow shear (TFS) profile on the dynamics of ITB evolution is investigated. It is found that the decorrelation between meso-scale modes and ITG driven modes due to the TFS can prevent global relaxation.

The mechanism of transport barrier collapse in reversed shear configuration is investigated using a global ion temperature gradient driven drift wave (ITG) turbulence simulation code. The transport barrier collapse event is explained in the context of non-local meso-scale interaction. It is found that the turbulence evolution and the sustainment of the transport barrier is sensitive to the safety factor profile. However, some common meso-scale modes are excited via various non-linear interaction channels whether or not a meso-scale mode in the vicinity of the q profile minimum has resonant surfaces. The non-linear interaction among these non-local modes makes long range correlation possible and carries energy from the core region to the q minimum region. It was shown that such linkage can lead to the collapse of the transport barrier and the change of the global temperature profile.

The key points in our method of repair of cryptotia are (1) to cover the skin defect on the posterior aspect of the ear after it is dissected from the head, and (2) to repair the cartilaginous deformity. We describe the use of a temporal skin flap for the former. For the latter, we make parallel incisions on the back of the superior crus and transfer a small graft of conchal cartilage there. The conchal cartilage graft "splints" the repaired crus against cicatricial contracture.

Power exhaust scenario and divertor design for a steady-state Japan (JA) DEMO and a pulse Europe (EU) DEMO1 have been investigated as one of the most important common issues in Broader Approach DEMO Design Activity. Radiative cooling is a common approach for the power exhaust scenario. For the JA DEMO, development of the divertor design appropriate for high P-sep/R-p similar to 30 MW m(-1) is required, while the radiation fraction in the main plasma (f(rad)(main) = P-rad(main)/P-heat) is ITER-level (0.40-0.45) and the exhaust power above the L- to H-mode power threshold (f(LH) = P-sep/P-th(LH)) is large margin (similar to 2). For the EU DEMO1, larger f(rad)(main) (=0.67) and smaller f(LH) (= 1.2) plasma is required, using higher-Z impurity seeding, in order to apply ITER-level divertor (P-sep/R-P = 17 MW m(-1)). ITER technology, i.e. water cooling with W-monoblock and Cu-alloy (CuCrZr) heat sink, is a baseline for JA and EU to handle the peak heat load of 10 MW m(-2)-level, and neutron flux and irradiation dose are comparable. For the JA DEMO, two different water-cooling pipes, i.e. CuCrZr and F82H steel, are proposed. For the EU DEMO1, the heat sink consists of all Cu-alloy pipe, and the divertor size is reduced with replacing the baffles by the breeding blankets. Choices of the heat sink components have been developed appropriate to the high irradiation dose condition. These JA and EU approaches of the power exhaust scenario will provide important case studies for the future decision of the DEMO divertor design.

This paper summarizes the evolution of Japanese DEMO design studies in a retrospective manner by highlighting efforts to resolve critical design issues on DEMO. Japan is currently working on the conceptual study of a steady-state DEMO (JA DEMO) with a major radius Rp of 8.5 m and fusion power Pfus of 1.5 to 2 GW based on water-cooled solid breeding blanket with pressurized water reactor water condition (290ºC to 325ºC, 15.5 MPa). Such a lower Pfus allows to find realistic design solutions for divertor heat removal. Recognizing that divertor heat removal is one of the most challenging issues on DEMO, the divertor design has been carried out in different approaches, including numerical divertor plasma simulation, magnetic configurations, heat sink design, etc. It is noteworthy that the latest divertor simulation led to a design window allowing divertor heat removal of the peak heat flux of

We report an additional surgical method to form a philtrum in cleft lip cases. The key point of this new method is to release the skin tension of the upper lip, thus protecting against the disappearance of the newly-made philtrum. We have obtained better results with this method, in comparison with the other methods previously reported.