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[en] We use the guiding center code ORBIT to study the broadening of resonances and the parametric dependence of the resonance frequency broadening width ΔΩ on the nonlinear particle trapping frequency ωb of wave-particle interaction with specific examples using realistic equilibrium DIII-D shot 159243 (Collins et al. 2016 Phys. Rev. Lett. 116 095001). When the mode amplitude is small, the pendulum approximation for energetic particle dynamics near the resonance is found to be applicable and the ratio of the resonance frequency width to the deeply trapped bounce frequency ΔΩ/ωb equals 4, as predicted by theory. Lastly, it is found that as the mode amplitude increases, the coefficient a=ΔΩ/ωb becomes increasingly smaller because of the breaking down of the nonlinear pendulum approximation for the wave-particle interaction.
[en] The creation and subsequent evolution of marginally-unstable modes have been observed in a wide range of fusion devices. This behaviour has been successfully explained, for a single frequency shifting mode, in terms of phase-space structures known as a 'hole' and 'clump'. Here in this paper, we introduce stochasticity into a 1D kinetic model, affecting the formation and evolution of resonant modes in the system. We find that noise in the fast particle distribution or electric field leads to a shift in the asymptotic behaviour of a chirping resonant mode; this noise heuristically maps onto radial microturbulence via canonical toroidal momentum scattering, affecting hole and clump formation. While the mechanism allowing for the formation of the hole and clump is coherent, the lifetime of a hole and clump is shown to be highly sensitive to initial conditions, affecting the temporal profile of a single bursting event in mode amplitude.
[en] Experiments conducted in high-performance 1.0 MA and 1.2 MA 6 MW NBI-heated H-mode discharges with a high magnetic flux expansion radiative divertor in NSTX demonstrate that significant divertor peak heat flux reduction and access to detachment may be facilitated naturally in a highly-shaped spherical torus (ST) configuration. Improved plasma performance with high βt = 15-25%, a high bootstrap current fraction fBS = 45-50%, longer plasma pulses, and an H-mode regime with smaller ELMs has been achieved in the strongly-shaped lower single null configuration with elongation κ = 2.2-2.4 and triangularity (delta) = 0.6-0.8. Divertor peak heat fluxes were reduced from 6-12 MW/m2 to 0.5-2 MW/m2 in ELMy H-mode discharges using the inherently high magnetic flux expansion fm = 16-25 and the partial detachment of the outer strike point at several D2 injection rates. A good core confinement and pedestal characteristics were maintained, while the core carbon concentration and the associated Zeff were reduced. The partially detached divertor regime was characterized by an increase in divertor radiated power, a reduction of ion flux to the plate, and a large neutral compression ratio. Spectroscopic measurements indicated a formation of a high-density, low temperature region adjacent to the outer strike point, where substantial increases in the volume recombination rate and CII, CIII emission rates was measured
[en] Here, after completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3 × 1019 m–3, central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre.
[en] Halo currents generated during unmitigated tokamak disruptions are known to develop rotating asymmetric features that are of great concern to ITER because they can dynamically amplify the mechanical stresses on the machine. This paper presents a multi-machine analysis of these phenomena. More specifically, data from C-Mod, NSTX, ASDEX Upgrade, DIII-D, and JET are used to develop empirical scalings of three key quantities: the machine-specific minimum current quench time, τCQ; the halo current rotation duration, trot; and (3) the average halo current rotation frequency, 〈fh〉 . These data reveal that the normalized rotation duration, trot/τCQ , and the average rotation velocity, 〈vh〉 , are surprisingly consistent from machine to machine. Furthermore, comparisons between carbon and metal wall machines show that metal walls have minimal impact on the behavior of rotating halo currents. Finally, upon projecting to ITER, the empirical scalings indicate that substantial halo current rotation above 〈 fh〉=20 Hz is to be expected. More importantly, depending on the projected value of τCQ in ITER, substantial rotation could also occur in the resonant frequency range of 6–20 Hz. In conclusion, the possibility of damaging halo current rotation during unmitigated disruptions in ITER cannot be ruled out.
[en] Here, a technique for tokamak equilibrium reconstructions is used for multiple DIII-D discharges, including L-mode and H-mode cases when weakly 3D fields (δB/B∼10"−"3) are applied. The technique couples diagnostics to the non-linear, ideal MHD equilibrium solver VMEC, using the V3FIT code, to find the most likely 3D equilibrium based on a suite of measurements. It is demonstrated that V3FIT can be used to find non-linear 3D equilibria that are consistent with experimental measurements of the plasma response to very weak 3D perturbations, as well as with 2D profile measurements. Observations at DIII-D show that plasma rotation larger than 20 krad s"–"1 changes the relative phase between the applied 3D fields and the measured plasma response. Discharges with low averaged rotation (10 krad s"–"1) and peaked rotation profiles (40 krad s"–"1) are reconstructed. Similarities and differences to forward modeled VMEC equilibria, which do not include rotational effects, are shown. Toroidal phase shifts of up to 30"∘ are found between the measured and forward modeled plasma responses at the highest values of rotation. The plasma response phases of reconstructed equilibra on the other hand match the measured ones. This is the first time V3FIT has been used to reconstruct weakly 3D tokamak equilibria.
[en] Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasma and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. In conclusion, the SFD resulted in a 2.5×reduction in the peak heat flux for many energy confinement times (2–3s) without any adverse effects on core plasma performance.
[en] Here, a recent DIII-D experiment investigating the impact of electron cyclotron heating (ECH) on neutral beam driven reversed shear Alfvén eigenmode (RSAE) activity is presented. The experiment includes variations of ECH injection location and timing, current ramp rate, beam injection geometry (on/off-axis), and neutral beam power. Essentially all variations carried out in this experiment were observed to change the impact of ECH on AE activity significantly. In some cases, RSAEs were observed to be enhanced with ECH near the off-axis minimum in magnetic safety factor (q_m_i_n), in contrast to the original DIII-D experiments where the modes were absent when ECH was deposited near q_m_i_n . It is found that during intervals when the geodesic acoustic mode (GAM) frequency at q_m_i_n is elevated and the calculated RSAE minimum frequency, including contributions from thermal plasma gradients, is very near or above the nominal TAE frequency (f_T_A_E), RSAE activity is not observed or RSAEs with a much reduced frequency sweep range are found. This condition is primarily brought about by ECH modification of the local electron temperature (T_e) which can raise both the local T_e at q_m_i_n as well as its gradient. A q-evolution model that incorporates this reduction in RSAE frequency sweep range is in agreement with the observed spectra and appears to capture the relative balance of TAE or RSAE-like modes throughout the current ramp phase of over 38 DIII-D discharges. Detailed ideal MHD calculations using the NOVA code show both modification of plasma pressure and pressure gradient at q_m_i_n play an important role in modifying the RSAE activity. Analysis of the ECH injection near the q_m_i_n case where no frequency sweeping RSAEs are observed shows the typical RSAE is no longer an eigenmode of the system. What remains is an eigenmode with poloidal harmonic content reminiscent of the standard RSAE, but absent of the typical frequency sweeping behavior. The remaining eigenmode is also often strongly coupled to gap TAEs. Analysis with the non-perturbative gyro fluid code TAEFL confirms this change in RSAE activity and also shows a large drop in the resultant mode growth rates.
[en] The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX) – a tokamak specifically designed to address critical gaps in the world fusion research program on the pathway to next step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP), and/or demonstration power plant (DEMO). This high field (≥ 6.5 tesla, 1.5 MA), high power density facility (P/S ~ 1.5 MW/m2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters – reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region – while simultaneously producing high performance core plasma conditions that are prototypical of a reactor: equilibrated electrons and ions, regimes with low or no torque, and no fueling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators with the unprecedented ability to deploy launch structures both on the lowmagnetic- field side and the high-magnetic-field side – the latter being a location where energetic plasma-material interactions can be controlled and favorable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination – advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions – will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady state sustainment would pave the way toward an attractive pilot plant, as envisioned in the ARC concept (Affordable, Robust, Compact) [B. N. Sorbom, et al., submitted to Fusion Engineering Design, 2014] that makes use of high-temperature superconductor technology – a high-field (9.25 tesla) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.
[en] Initial tests on the National Spherical Torus Experiment Upgrade (NSTX-U) (Menard et al 2012 Nucl. Fusion 083015) device suggest that introducing energy selectivity for sawtooth induced fast ion redistribution is required to improve the agreement between experimental and simulated quantities such as neutron rate and Fast-Ion D-Alpha profiles. The aim of this work is to assess the requirements to properly describe the behaviour of fast ions during a sawtooth crash for predictive sawtooth simulations. As the first step, in this work, we use the particle-following Orbit code to characterize the redistribution of fast particles. In order for a sawtooth crash to be simulated, a spatial and temporal displacement is implemented into the Orbit code. The perturbation amplitude is determined by comparison with experimental measurement of the neutron rate drop. The characteristics of fast ions with different orbit types are investigated in phase and real space. Due to a sawtooth crash, fast ion energy and angular momentum are modified resulting in the redistribution in phase space and orbit type change. The redistribution of fast ions in real space shows that the sawtooth instability brings different effect on fast particles with different orbit types as observed in experiments. Finally, the initial interpretative Transp simulation using the so-called kick model based on the Orbit modeling result shows an improvement of fast ion redistribution before and after a sawtooth crash but the neutron rate still has discrepancy compared to the experimental measurement.