The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

/pdf/plasma-material-interactions-in-current-tokamaks-and-their-1a6iwv89x6.pdf

Plasma{material interactions in current tokamaks and their implications for next step fusion reactors

Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

https://iopscience.iop.org/article/10.1088/0029-5515/47/6/S03/pdf

Chapter 3: MHD stability, operational limits and disruptions

The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions. Very considerable progress has been made in understanding, controlling and predicting tokamak transport across a wide variety of plasma conditions and regimes since the publication of the ITER Physics Basis (IPB) document (1999 Nucl. Fusion 39 2137-2664). Major areas of progress considered here follow. (1) Substantial improvement in the physics content, capability and reliability of transport simulation and modelling codes, leading to much increased theory/experiment interaction as these codes are increasingly used to interpret and predict experiment. (2) Remarkable progress has been made in developing and understanding regimes of improved core confinement. Internal transport barriers and other forms of reduced core transport are now routinely obtained in all the leading tokamak devices worldwide. (3) The importance of controlling the H-mode edge pedestal is now generally recognized. Substantial progress has been made in extending high confinement H-mode operation to the Greenwald density, the demonstration of Type I ELM mitigation and control techniques and systematic explanation of Type I ELM stability. Theory-based predictive capability has also shown progress by integrating the plasma and neutral transport with MHD stability. (4) Transport projections to ITER are now made using three complementary approaches: empirical or global scaling, theory-based transport modelling and dimensionless parameter scaling (previously, empirical scaling was the dominant approach). For the ITER base case or the reference scenario of conventional ELMy H-mode operation, all three techniques predict that ITER will have sufficient confinement to meet its design target of Q = 10 operation, within similar uncertainties.

https://iopscience.iop.org/article/10.1088/0029-5515/47/6/S02/pdf

Chapter 2: Plasma confinement and transport

Quantum phase transitions and vortex dynamics in superconducting networks

In magnetic fusion reactors relying on the burning of deuterium and tritium, sufficient confinement of the alpha particles produced in the nuclear reactions is crucial to sustaining the burning plasma. In this article the interactions of these energetic particles with linear and nonlinear Alfve'n waves generated in the magnetized plasma are reviewed.

https://link.aps.org/accepted/10.1103/RevModPhys.88.015008

Physics of Alfvén waves and energetic particles in burning plasmas

Development of 170 and 110 GHz gyrotrons and their application to electron cyclotron heating systems are presented. A parasitic oscillation that degraded the electron beam quality was suppressed, and the performance of the gyrotron improved significantly. Up to now, 0.9 MW/9.2 s, 0.5 MW/30 s, 0.3 MW/60 s, 0.2 MW/132 s, etc, have been demonstrated at 170 GHz. At 110 GHz, 1.3 MW/1.5 s, 1.2 MW/4.1 s, 1 MW/5 s were obtained. It is found that the reduction of stray radiation and the enhancement of cooling capability are important for continuous wave operation. Four 110 GHz gyrotrons are under operation in the electron cyclotron heating and current drive system of JT-60U. Power up to approximately 3 MW/2.7 s was injected into the plasma through movable mirrors in the poloidal direction and contributed to electron heating and the suppression of the neo-classical tearing modes.

https://iopscience.iop.org/article/10.1088/0029-5515/43/8/314/pdf

Development of 170 and 110 GHz gyrotrons for fusion devices

Interactions between fusion materials R&D and other technologies

Electron cyclotron heating (ECH)-assisted startup experiments have been performed in JT-60U. The breakdown loop voltage was successfully reduced from 25 to 4 V (=0.26 V m−1) by 200 kW ECH. This is lower than the 0.3 V m−1, which corresponds to the maximum electric field in ITER. Parameter scans of ECH power, prefill pressure, resonance position and polarization were carried out. The sensitivity of the breakdown to polarization and resonance position was observed. A prefilling gas pressure scan showed that the initial breakdown density increases with prefill pressure when it is is lower than 8 × 10−5 Torr. Higher harmonic ECH was also attempted. The second harmonic ECH-assisted startup was possible with higher ECH power injection. However, the third harmonic ECH-assisted startup was not successful.

https://iopscience.iop.org/article/10.1088/0029-5515/45/7/019/pdf

Electron cyclotron heating assisted startup in JT-60U

In the JT-60U tokamak a pre-formed reversed magnetic shear configuration was sustained non-inductively by means of lower hybrid current drive (LHCD). It was formed by neutral beam heating at current build-up and had a reversed shear region extending to 65% of the minor radius. The configuration was sustained for 7.5 s, and the extent of the reversed shear region was kept as large as 55% of the minor radius until the very end. The plasma was stable against MHD activities. It is also demonstrated that LHCD could modify reversed magnetic shear keeping plasma parameters, such as the plasma current and the toroidal magnetic field, the same.

https://iopscience.iop.org/article/10.1088/0741-3335/38/10/001/pdf

Sustainment and modification of reversed magnetic shear by LHCD on JT-60U

In JT-60U lower hybrid current drive (LHCD) experiments, a reversed magnetic shear configuration that was accompanied by internal transport barriers was successfully maintained in a quasi-steady state almost entirely non-inductively by the LH driven current and the bootstrap current. The normalized β was kept near 1 and the neutron emission rate was nearly steady as well, indicating no accumulation of impurities in the plasma. Diagnostics showed that the profiles of the electron and ion temperatures, the electron density and the current were almost unchanged during the LHCD phase. The applicability of LHCD to H mode plasmas has also been investigated. It was found that the LH waves can be coupled to an H mode edge plasma with a plasma-wall distance of up to about 14 cm.

/pdf/lhcd-current-profile-control-experiments-towards-steady-4z74zqceji.pdf

Masami Seki

Papers

Development of 170 and 110 GHz gyrotrons for fusion devices

Interactions between fusion materials R&D and other technologies

Electron cyclotron heating assisted startup in JT-60U

Sustainment and modification of reversed magnetic shear by LHCD on JT-60U

LHCD current profile control experiments towards steady state improved confinement on JT-60U