論文

© 2018 Elsevier B.V. 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 ∼30 MW m −1 is required, while the radiation fraction in the main plasma (f radmain = P radmain /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 thLH ) is large margin (∼2). For the EU DEMO1, larger f radmain (=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.

© 2018 Elsevier B.V. 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.

© 2018 IAEA, Vienna. The integrated modelling code TOPICS has been extended to include core impurity transport, and applied to predictive modelling of JT-60SA high-beta steady-state plasma with the accumulation of impurity seeded to reduce the divertor heat load. In the modelling, models and conditions are selected for a conservative prediction, which considers a lower bound of plasma performance with the maximum accumulation of impurity. The conservative prediction shows the compatibility of impurity seeding with core plasma with high-beta (β N > 3.5) and full current drive conditions, i.e. when Ar seeding reduces the divertor heat load below 10 MW m-2, its accumulation in the core is so moderate that the core plasma performance can be recovered by additional heating within the machine capability to compensate for Ar radiation. Due to the strong dependence of accumulation on the pedestal density gradient, high separatrix density is important for the low accumulation as well as the low divertor heat load. The conservative prediction also shows that JT-60SA has enough capability to explore the divertor heat load control by impurity seeding in high-beta steady-state plasmas.

© 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim The mirror effect on scrape-off layer (SOL)-divertor plasma profiles was studied by relaxing the assumption of homogeneous magnetic fields used in a one-dimensional (1D) plasma fluid model, which incorporates the anisotropic ion temperature. For collisionless plasmas, the perpendicular ion temperature proportional to the magnetic field strength is observed, which reflects the conservation of the magnetic moment. In diverging field cases, the plasma flow becomes supersonic, while it is kept subsonic in contracting field cases. Parallel momentum gains/losses and energy transfers between parallel and perpendicular components caused by the mirror effect are also observed. In addition, it is indicated that the viscous flux approximation is invalid in inhomogeneous magnetic fields regardless of the plasma collisionality.

© 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim We have developed an extended kinetic model for the thermal force (Formula presented.) on test impurity particles in plasmas. This model makes use of a combination of the plasma ion heat flux q GM derived from the generalized moment equations (E. Zawaideh et al., Phys. Fluids 1988, 31) and the numerical algorithm presented earlier (Y. Homma, A. Hatayama, J. Comput. Phys. 2013, 250). The applicable condition extends from collisional to weakly collisional plasmas, and (Formula presented.) weakens when the plasma collisionality decreases. A test simulation is performed assuming a tentative weakly collisional plasma. The results show that it can have a non-negligible influence on transient impurity transport in a DEMO-scale tokamak to take into account the effect of plasma collisionality in thermal force modelling.

© 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim The effects of multi-impurity transport are important for the divertor plasma performance. To simulate the radiative divertor plasma precisely, we have extended the integrated divertor code, SONIC, to the simulation with multi-impurity species. For SONIC extension, a new integrated modelling framework using the multiple-program multiple-data approach and message passing interface data exchange has been developed. In the framework, two impurities and more can be treated simultaneously with several Monte Carlo kinetic impurity codes. The extended SONIC code has been applied to analyse the JT-60SA radiative divertor plasma. The SONIC simulation including multi-impurity transport demonstrates that the radiative divertor plasma scenario compatible with the steady-state high-β plasma can be achieved.

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%.

Scrape Off Layer (SOL) plasma circumnavigates the torus along magnetic field lines, and some part of it encounters a leading edge of the first wall (FW). This leads the concentration of the plasma energy in a small area, causing a hot spot. In a DEMO reactor that deals with a few hundreds of MW of plasma energy in the SOL, the leading edge problem is a critical issue in the FW design. For the design purposes in DEMO and power reactors, a new analysis method is introduced, where a flux tube is poloidally divided into the Apple Peel Like Elements (APPLEs) and each APPLE is enclosed by adjacent 4 magnetic field lines. Considering the contact area of the APPLEs on the FW and the radial energy transport between the APPLEs, FW heat load profile can be analyzed. The result of the case that the fusion power is 1.5 GW, and the major radius is 8.2 m, shows that the heat load is peaked to about 1 MW/m2 near the inboard midplane and the baffle plate. This method is expected to be useful for the DEMO design.

Power exhaust to the divertor and the conceptual design have been investigated for a steadystate 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 Psep = 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 Pout= 250 MW and the total radiation fraction at the edge, SOL and divertor (Prad/ Pout = 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 (qtarget) 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 qtarget 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 qtarget 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.

Erosion reduction of tungsten (W) divertor target is one of the most important research subjects for the DEMO fusion reactor design, because the divertor target has to sustain large fluence of incident particles, composed mainly of fuel ions and seeded impurities, during year-long operation period. Rate of net erosion and deposition on outer divertor target has been studied by using the integrated SOL/divertor plasma code SONIC and the kinetic full-orbit impurity transport code IMPGYRO. Two background plasmas have been used: one is lower density n i and higher temperature case and the other is higher n i and lower temperature case. Net erosion has been seen in the lower n i case. But in the higher n i case, the net erosion has been almost suppressed due to increased return rate and reduced self-sputtering yield. Following two factors are important to understand the net erosion formation: (i) ratio of the 1st ionization length of sputtered W atom to the Larmor gyro radius of W+ ion, (ii) balance between the friction force and the thermal force exerted on W ions. DEMO divertor design should take into account these factors to prevent target erosion.

An extended Computed Tomography (CT) technique with a priori information based on Phillips-Tikhonov regularization has been developed to handle a high-contrast radiation distribution, which can result in large reconstruction errors in the region where radiation intensity is low, with few line of sight (LOS) data. Reference profiles generated from LOS data for every time slice are employed as the a priori information. In the extended technique, the weighting parameter for the reference profile is automatically determined from the LOS data, to avoid an inappropriate reference. The extended technique has been examined with a reconstruction test with the previously designed CT system using an infrared imaging video bolometer in JT-60SA. In the reconstructed profile of the high-contrast radiation distribution, the extended technique shows improvement of the weak radiation region, such as the scrape-off layer, without the deterioration of the strong radiation region, such as around the divertor. The results indicate that the extended CT technique can help to handle a high-contrast radiation distribution with few LOS data.

In order to obtain a better understanding of tungsten (W) transport processes, we are developing the Monte-Carlo W transport code IMPGYRO. The code has the following characteristics which are important for calculating W transport: (1) the exact Larmor motion of W ions is computed so that the effects of drifts are automatically taken into account; (2) Coulomb collisions between W impurities and background plasma ions are modelled using the Binary Collision Model which provides more precise kinetic calculations of the friction and thermal forces. By using the IMPGYRO code, the W production/transport in the ITER geometry has been calculated under two different divertor operation modes (Case A: partially detached state and Case B: high recycling state) obtained from the SOLPS-ITER code suite calculation without the effect of drifts. The results of the W-density in the upstream SOL (scrape-off layer) strongly depend on the divertor operation mode. From the comparison of the W impurity transport between Case A and Case B, obtaining a partially detached state is shown to be effective to reduce W-impurities in the upstream SOL. The limitations of the employed model and the validity of the above results are discussed and future problems are summarized for further applications of IMPGYRO code to ITER plasmas.

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised.

An InfraRed imaging Video Bolometer (IRVB) system in JT-60SA has been designed for the two-dimensional (2D) Computed Tomography (CT) analysis of radiation. To achieve complete viewing of the whole poloidal cross section for plasmas having a low aspect ratio, a new IRVB concept, which has two sets of pin-hole camera systems viewing two different directions, has been introduced. Fields of view of the IRVB have been successfully designed by moving apertures to be sufficient for CT use without installation in tangential ports. The size of the aperture, which determines the width of the sightline of the IRVB channels, has also been optimized with respect to the reconstruction error with well-established solvers of tomography. The optimization indicates that the 7 mm square aperture is the best for this system, and the CT measurement without this optimization can become degraded. A synthetic image using an estimated radiation profile shows that the signal to noise (S/N) ratio of the designed IRVB is large enough to identify the incident radiation power. The result indicates that the designed IRVB can be used for the 2D CT measurement of radiation in JT-60SA.

The approximate estimation of tritium permeation rate under the acceptable assumption from a safety point of view is surely useful to progress the design activities for a fusion DEMO reactor. Tritium permeation rates in the blanket and the divertor were estimated by the simplified evaluation model under the recent DEMO conditions in the water-cooled blanket with solid breeder as a first step. Plasma driven permeation rates in tungsten wall were calculated by applying Doyle & Brice model and gas driven permeation rates in F82H were calculated for hydrogen-tritium twocomponent system. In the representative recent DEMO condition, the following tritium permeation rates were obtained, 1.8 g/day in the blanket first wall, 2.3 g/day in the blanket tritium breeding region and 1.6 g/day in the divertor. Total tritium permeation rate into the cooling water was estimated to be 5.7 g/day.