A future fusion reactor is expected to have all-metal plasma facing materials (PFMs) to ensure low erosion rates, low tritium retention and stability against high neutron fluences. As a consequence, intrinsic radiation losses in the plasma edge and divertor are low in comparison to devices with carbon PFMs. To avoid localized overheating in the divertor, intrinsic low-Z and medium-Z impurities have to be inserted into the plasma to convert a major part of the power flux into radiation and to facilitate partial divertor detachment. For burning plasma conditions in ITER, which operates not far above the L–H threshold power, a high divertor radiation level will be mandatory to avoid thermal overload of divertor components. Moreover, in a prototype reactor, DEMO, a high main plasma radiation level will be required in addition for dissipation of the much higher alpha heating power. For divertor plasma conditions in present day tokamaks and in ITER, nitrogen appears most suitable regarding its radiative characteristics. If elevated main chamber radiation is desired as well, argon is the best candidate for the simultaneous enhancement of core and divertor radiation, provided sufficient divertor compression can be obtained. The parameter Psep/R, the power flux through the separatrix normalized by the major radius, is suggested as a suitable scaling (for a given electron density) for the extrapolation of present day divertor conditions to larger devices. The scaling for main chamber radiation from small to large devices has a higher, more favourable dependence of about Prad,main/R2. Krypton provides the smallest fuel dilution for DEMO conditions, but has a more centrally peaked radiation profile compared to argon. For investigation of the different effects of main chamber and divertor radiation and for optimization of their distribution, a double radiative feedback system has been implemented in ASDEX Upgrade (AUG). About half the ITER/DEMO values of Psep/R have been achieved so far, and close to DEMO values of Prad,main/R2, albeit at lower Psep/R. Further increase of this parameter may be achieved by increasing the neutral pressure or improving the divertor geometry.

/pdf/impurity-seeding-for-tokamak-power-exhaust-from-present-1u11q729cm.pdf

Impurity seeding for tokamak power exhaust: from present devices via ITER to DEMO

Feedback control of the divertor power load by means of nitrogen seeding has been developed into a routine operational tool in the all-tungsten clad ASDEX Upgrade tokamak. For heating powers above about 12?MW, its use has become inevitable to protect the divertor tungsten coating under boronized conditions. The use of nitrogen seeding is accompanied by improved energy confinement due to higher core plasma temperatures, which more than compensates the negative effect of plasma dilution by nitrogen on the neutron rate. This paper describes the technical details of the feedback controller. A simple model for its underlying physics allows the prediction of its behaviour and the optimization of the feedback gain coefficients used. Storage and release of nitrogen in tungsten surfaces were found to have substantial impact on the behaviour of the seeded plasma, resulting in increased nitrogen consumption with unloaded walls and a latency of nitrogen release over several discharges after its injection. Nitrogen is released from tungsten plasma facing components with moderate surface temperature in a sputtering-like process; therefore no uncontrolled excursions of the nitrogen wall release are observed. Overall, very stable operation of the high-Z tokamak is possible with nitrogen seeding, where core radiative losses are avoided due to its low atomic charge Z and a high ELM frequency is maintained.

/pdf/divertor-power-load-feedback-with-nitrogen-seeding-in-asdex-3ahu08lbrk.pdf

Divertor power load feedback with nitrogen seeding in ASDEX Upgrade

Plasma surface interactions in impurity seeded plasmas

A Korean fusion energy development promotion law (FEDPL) was enacted in 2007. As a following step, a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012. After the thorough 0D system analysis, the parameters of the main machine characterized by the major and minor radii of 6.8 and 2.1 m, respectively, were chosen for further study. The analyses of heating and current drives were performed for the development of the plasma operation scenarios. Preliminary results on lower hybrid and neutral beam current drive are included herein. A high performance Nb3Sn-based superconducting conductor is adopted, providing a peak magnetic field approaching 16 T with the magnetic field at the plasma centre above 7 T. Pressurized water is the prominent choice for the main coolant of K-DEMO when the balance of plant development details is considered. The blanket system adopts a ceramic pebble type breeder. Considering plasma performance, a double-null divertor is the reference configuration choice of K-DEMO. For a high availability operation, K-DEMO incorporates a design with vertical maintenance. A design concept for K-DEMO is presented together with the preliminary design parameters.

https://iopscience.iop.org/article/10.1088/0029-5515/55/5/053027/pdf

Design concept of K-DEMO for near-term implementation

The JT-60SA project implemented by Japan and Europe is progressing on schedule towards the first plasma in March 2019. After careful R&D, procurements of the major components have entered their manufacturing stages. In parallel, disassembly of JT-60U has been completed on time, and the JT-60SA tokamak assembly is expected to start in January 2013. The JT-60SA device, a highly shaped large superconducting tokamak with a variety of plasma control actuators, has been designed in order to contribute to ITER and to complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. Detailed assessments and prediction studies of the JT-60SA plasma regimes have confirmed these capabilities: using ITER- and DEMO-relevant plasma regimes, heating conditions, and its sufficiently long discharge duration, JT-60SA enables studies on magnetohydrodynamic stability at high beta, heat/particle/momentum transport, high-energy ion physics, pedestal physics including edge localized mode control, and divertor physics. By integrating these studies, the project provides ‘simultaneous and steady-state sustainment of the key performance characteristics required for DEMO’ with integrated control scenario development.

https://iopscience.iop.org/article/10.1088/1741-4326/ab03f6/pdf

Progress of the JT-60SA project

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

https://iopscience.iop.org/article/10.1088/0029-5515/49/7/075029/pdf

Compact DEMO, SlimCS: design progress and issues

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Compact DEMO, SlimCS: design progress and issues