The scan of Ion Cyclotron Resonant Heating power has been used to systematically study the pump out effect of central electron heating on impurities such as Ni and Mo in H mode low collisionality discharges in JET. The transport parameters of Ni and Mo have been measured by introducing a transient perturbation on their densities via the Laser Blow Off technique. Without ICRH, Ni and Mo density profiles are typically peaked. The application of ICRH, induces on Ni and Mo in the plasma center (at normalized poloidal flux r = 0.2) an outward drift approximately proportional to the amount of injected power. Above a threshold, of about 3MW of ICRH power in the specific case, the radial flow of Ni and Mo changes from inward to outward and the impurity profiles, extrapolated to stationary conditions, become hollow. At mid radius the impurity profiles become flat or only slightly hollow. In the plasma centre the variation of the pinch parameter v/D of Ni is particularly well correlated with the change of the ion temperature gradient, in qualitative agreement with the neoclassical theory. However, the experimental radial velocity is larger than the neoclassical one by up to one order of magnitude. Gyrokinetic simulations of the radial impurity fluxes induced by electrostatic turbulence do not foresee a flow reversal in the analyzed discharges.

The next step in the Wendelstein stellarator line is the large superconducting device Wendelstein 7-X, currently under construction in Greifswald, Germany. Steady-state operation is an intrinsic feature of stellarators, and one key element of the Wendelstein 7-X mission is to demonstrate steady-state operation under plasma conditions relevant for a fusion power plant. Steady-state operation of a fusion device, on the one hand, requires the implementation of special technologies, giving rise to technical challenges during the design, fabrication and assembly of such a device. On the other hand, also the physics development of steady-state operation at high plasma performance poses a challenge and careful preparation. The electron cyclotron resonance heating system, diagnostics, experiment control and data acquisition are prepared for plasma operation lasting 30 min. This requires many new technological approaches for plasma heating and diagnostics as well as new concepts for experiment control and data acquisition.

The JET 2019-2020 scientific and technological programme exploited the results of years of concerted scientific and engineering work, including the ITER-like wall (ILW: Be wall and W divertor) installed in 2010, improved diagnostic capabilities now fully available, a major Neutral Beam Injection (NBI) upgrade providing record power in 2019-2020, and tested the technical & procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power and alpha particle physics in the coming D-T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation. The efficacy of the newly installed Shattered Pellet Injector for mitigating disruption forces and runaway electrons was demonstrated. Secondly, research on the consequences of long-term exposure to JET-ILW plasma was completed, with emphasis on wall damage and fuel retention, and with analyses of wall materials and dust particles that will help validate assumptions and codes for design & operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver maximum technological return from operations in D, T and D-T benefited from the highest D-D neutron yield in years, securing results for validating radiation transport and activation codes, and nuclear data for ITER.

Most present Ion Cyclotron Resonant Frequency (ICRF) heating codes and antenna codes assume the antenna sitting in a vacuum region and consider the fast wave only, which implicitly performs an abrupt density transition from vacuum to above lower hybrid (LH) resonance. The impact of the appearance of the LH resonance is entirely overlooked in their simulations. We studied the impact of densities that decay continuously inside the antenna box on near field patterns and power coupling. A new full wave code based on the COMSOL Finite Element Solver has been developed to investigate this topic. It is shown that: up to the memory limits of the adopted workstation, the local RF field pattern in low-density regions below the LH resonance changes with the grid size. Interestingly and importantly, however, the total coupled toroidal spectrum is almost independent on the mesh size and is weakly affected by the presence of the density profile inside the antenna box in dipole toroidal strap phasing. This suggests one can drop out this density for coupling studies to speed up the computation. Simulation also shows that varying the density gradient in the fast wave evanescence region has no significant effect on wave coupling.

Achieving high neutron yields in today’s fusion research relies on high-power auxiliary heating in order to attain required core temperatures. This is usually achieved by means of high neutral beam (NB) and radio frequency (RF) power. Application of NB power is accompanied by production of fast beam ions and associated beam–target (BT) reactions. In standard JET operational conditions, deuterium (D) NBs are injected into D plasmas. The injected beams comprise D atoms at full, one-half and one-third injected energy. Typically, the full energy of the injected D beams is between 90 and 120 keV, providing 1.4–2.0 MW of heating, which is about half of the injected power. Half-energy D beams carry about one-third of the injected power and the rest of the power is carried by the third energy fraction of D beams. Under these conditions, thermal fusion reactions, i.e. those between plasma ions, and BT reactions are of the same order of magnitude. This study addresses important issues regarding the impact of density, central electron and ion temperatures and their ratio, T i(0)/T e(0), on fusion performance, measured by the total neutron yield and BT neutron counts. NB/RF synergistic effects are discussed as well. It is demonstrated that thermal fusion gain increases linearly with normalised plasma pressure, β N, and confinement, B t τ. The BT neutrons are, however, more difficult to predict and this task in general requires numerical treatment. In this study, BT neutrons in JET’s best-performing baseline and hybrid pulses are analysed and the underlying dependencies discussed. Central fast ion densities are found to decrease with increased density and density peaking. This is attributed to poorer beam penetration at high density. The BT reactions however are unchanged and can even increase if operating at higher core temperatures. An increase in the central ion temperature and T i(0)/T e(0) ratio leads to higher total and BT reaction rates whilst simultaneously the ratio of the BT to total neutron decreases significantly. NB/RF synergistic effects are found to have a negligible impact on total neutron rate. This can be explained by the reduced beam penetration in high-density conditions leading to lower central fast ion density.

A steady-state tokamak reactor (SSTR) requires a high efficiency current drive system, from plug to driven mega-amps. RF systems working in the ion-cyclotron range of frequencies (ICRF) have high efficiency from plug to antenna but a limited current drive (CD) efficiency and centrally peaked CD profiles. The latter feature is not adequate for a SSTR where the current should be sufficiently broad to keep the central safety factor (possibly significantly) above 1. In addition, the fact that the fast wave (FW) is evanescent at the edge limits coupling, requiring high voltage operation, which makes the system dependent on plasma edge properties and prone to arcing, reducing its reliability. A possible way to overcome these weaknesses is to operate at higher frequency (10 times or more the cyclotron frequency). The advantages are: (1) The coupling can be much better (waves propagate in vacuum) if the parallel refractive index n // is kept below one, (2) The FW group velocity tends to align to the magnetic field, so the power circumnavigates the magnetic axis and can drive offaxis current, (3) Due to the latter property, n // can be upshifted along the wave propagation path, allowing low n // launch (hence good coupling, large CD efficiency) with ultimately good electron absorption (which requires higher n //). Note however that the n // upshift is a self-organized feature, that electron absorption is in competition with-particle absorption and that uncoupling of the FW from the lower hybrid resonance at the edge requires n // slightly above one. The latter possibly counterproductive features might complicate the picture. The different aspects of this potentially attractive off-axis FWCD scheme are discussed.

The isotopic exchange efficiencies of JET Ion Cyclotron Wall Conditioning (ICWC) discharges produced at ITER half and full field conditions are compared for JET carbon (C) and ITER like wall (ILW). Besides an improved isotope exchange rate on the ILW providing cleaner plasma faster, the main advantage compared to C-wall is a reduction of the ratio of retained discharge gas to removed fuel. Complementing experimental data with discharge modeling shows that long pulses with high (~240kW coupled) ICRF power maximizes the wall isotope removal per ICWC pulse. In the pressure range 1 to 7.5 x10 -3 Pa, this removal reduces with increasing discharge pressure. As most of the wall-released isotopes are evacuated by vacuum pumps in the post discharge phase, duty cycle optimization studies for ICWC on JET-ILW need further consideration. The accessible reservoir by H 2 -ICWC at ITER half field conditions on the JET-ILW preloaded by D 2 tokamak operation is estimated to be 7.3 x10 22 hydrogenic atoms, and may be exchanged within 400s of cumulated ICWC discharge time.

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