FY2019 Research Milestones

JRT-2019:  Conduct research to understand the role of neutral fueling and transport in determining the pedestal structure.

The edge pedestal is a key component in achieving overall high confinement in a magnetic fusion device. Therefore, obtaining a physics understanding and predictive capability for the pedestal height and structure is a major goal of fusion research and requires advances in the understanding of the separate structure of density and temperature profiles in the pedestal region. A key challenge is to understand the importance of particle sources in determining the density pedestal and project to burning plasma scenarios. Experiments on DIII-D and archived data from C-Mod, DIII-D, and NSTX will be used to test how fueling, reduced recycling, and transport affect the density pedestal structure. The role of divertor geometry and its effect upon the pedestal structure will also be investigated. U.S. researchers involved in collaborative activities on other relevant experiments may also contribute to this effort.

R(19-1):  Assess H-mode energy confinement and pedestal characteristics with higher field, plasma current, and NBI heating power

Description:  Future ST devices such as ST-FNSF will operate at higher toroidal field, plasma current and heating power than NSTX.  To establish the physics basis for future STs, which are generally expected to operate in lower collisionality regimes, it is important to characterize confinement and pedestal structure over an expanded range of engineering parameters.  H-mode studies in NSTX and MAST have shown that the global energy confinement exhibits a more favorable scaling with collisionality (BtE ~ 1/n*e) than that from ITER98y,2.  In addition, the H-mode pedestal pressure increases with ~IP2.  With higher BT, IP, and NBI power with beams at different tangency radii, NSTX-U and MAST-U provide an excellent opportunity to assess the core and boundary characteristics in regimes more relevant to future STs and to explore the accessibility to lower collisionality.  Specifically, the relation between H-mode energy confinement and pedestal structure with increasing IP, BT and PNBI will be determined and compared with previous NSTX and MAST results, including emphasis on the collisionality dependence of confinement and beta dependence of pedestal width. Coupled with low-k turbulence diagnostics and gyrokinetic simulations, the experiments will provide further evidence for the mechanisms underlying the observed confinement scaling and pedestal structure.  During FY2019, significant effort will be put toward profile and turbulence diagnostic commissioning for these experiments on NSTX-U, and if NSTX-U cannot support plasma operations during FY2019, emphasis will be placed on collaboration on MAST-U to support the core transport and pedestal structure research goals of this milestone. 

R(19-2): Demonstrate optimized ramp-up scenarios in spherical tokamaks

Description:  This milestone leverages the simulation capabilities developed as part of the 18-6 milestone to realize optimized ramp-up scenarios on NSTX-U and MAST-U.  This requires the continued development of the simulation frameworks and dedicated experiments on the two devices.  The reduced model framework (TOKSYS) will be extended to include the evolution of the q-profile, temperature, density and fast-ion pressure.  The reduced models for the kinetic profiles will be derived from transport models, such as TRANSP, and experimental results. These advances in the model will be leveraged to examine the resiliency of the ramp-up scenarios to expected experimental variations, such as the timing of the L-H transition and the temporary loss or delayed turn-on of a neutral beam.  This framework will allow the development of real-time algorithms that can improve the resiliency of the scenarios.  For example, experiments on NSTX-U demonstrated that a new control scheme that alters the outer boundary of the plasma shape to achieve a target inner gap distance improved the resiliency of the time of diverting to variations in the neutral beam heating. The transport modeling framework (TRANSP) will be extended to remove assumptions from the initial studies pursued in FY18. For example, the evolution of Te and Ti would be based on a flux-driven transport model that was benchmarked on a database of ST ramp-up results. This capability would be leveraged to examine the plasma resistivity and current profile relaxation, which are significant drivers of the internal inductance, ohmic flux consumption and MHD stability. Another goal would be to implement a neutral fueling model that would alter the edge fueling to match a target L-mode density, similar to what can be achieved in the experiment.  Targeted experiments on MAST-U and/or NSTX-U would be conducted to test the simulation results and provide continued refinement to the assumptions of the models.

R(19-3):  Validate tearing mode physics for tearing avoidance in high-performance scenarios

Description: Tearing modes (TMs) and neoclassical tearing modes (NTMs) can significantly limit the access to high-performance regimes in spherical tokamaks (STs) and standard-aspect-ratio advanced tokamaks (AT). Particularly in high βP and/or low-torque ITER baseline scenarios it is critical to predict the path to avoid or mitigate TM/NTMs throughout the entire discharge evolution including early ramp-up phase. The goal of this milestone is to first validate TM/NTM physics models and simulations in ST and AT regimes, and thereby to develop robust predictive capability and avoidance strategies of TM/NTMs. Existing experimental data from NSTX, FY-16 NSTX-U, and DIII-D will be investigated for validation, and predictive simulations will be performed for DIII-D and KSTAR advanced scenarios as well as MAST-U initial operations. Each of the key physics components influencing the onset of tearing instabilities will be systematically examined and compared with applicable codes. The effects by shaping including aspect ratio, and by equilibrium profiles and parameters including qmin will be extensively analyzed with resistive MHD codes such as resistive DCON or PEST-III on TM databases, and the effects by kinetic profiles including rotation, multi-species effects, or the interaction with 3D fields, will be studied with extended MHD codes such as M3D-C1 or MARS on selected TM/NTM datasets. The toroidally-generalized TM stability index in the simulations will be compared with the onset conditions observed in experiments, and will be tested for predictive TM avoidance. TM/NTM evolution with islands in experiments will also be analyzed, and compared with extended MHD codes to develop various stabilizing techniques including NTM entrainment. Any reduced TM/NTM physics models developed will be utilized for TM/NTM mitigation and avoidance in various future device operations including accelerated access to high-performance operating scenarios in NSTX-U.

R(19-4): Assess the effects of neutral beam injection parameters on the fast ion distribution function and neutral beam driven current profile

Description:  Accurate knowledge of neutral beam (NB) ion properties is of paramount importance for many areas of tokamak physics. NB ions modify the power balance, provide torque to drive plasma rotation and affect the behavior of MHD instabilities. Moreover, they determine the non-inductive NB driven current, which is crucial for future devices such as ITER, FNSF and STs with small or no central solenoid. With the additional more tangentially-aimed NB sources, NSTX-U is well equipped to characterize a broad parameter space of fast ion distribution (Fnb) and NB-driven current properties, with significant overlap with other STs such as MAST-U and conventional aspect ratio tokamaks such as DIII-D. The two main goals of this milestone are (i) to characterize the NB ion behavior and compare it with classical predictions, and (ii) to document the operating space of NB-driven current profile. If NSTX-U operations resume in FY19, Fnb will be characterized through the upgraded set of NSTX-U fast ion diagnostics (e.g. fast-ion D-alpha: FIDA, solid-state neutral particle analyzer: ssNPA, scintillator-based fast-lost-ion probe: sFLIP, neutron counters, and possibly a Fusion Products diagnostic) as a function of NB injection parameters (tangency radius, beam voltage) and magnetic field. Building on the initial results obtained in the NSTX-U FY-2016 run campaign, well controlled, single-source scenarios at low NB power will be used to compare fast ion behavior with classical models (e.g. the NUBEAM module of TRANSP) in the absence of fast ion driven instabilities. Collaborations with MAST-U and DIII-D are foreseen for joint studies on NB-CD and validation of the modeling tools. Diagnostics data will be interpreted through the “beam blip” analysis technique and other dedicated codes such as FIDASIM. Then, the NB-driven current profile will be documented for the NB parameter space attainable on the three devices, e.g. by comparing NUBEAM/TRANSP predictions to measurements from Motional Stark Effect (if available), complemented by vertical/tangential FIDA systems, ssNPA and neutron/fusion product diagnostics to assess modifications of the classically expected Fnb. Particular emphasis will be placed on documenting driven current profile variations as a function of injecting beam tangency radius. If NSTX-U cannot support plasma operations during FY2019, additional emphasis will be placed on collaboration on MAST-U and DIII-D to support the experimental research goals of this milestone on characterization of the fast ion distribution from NBI and of the NB-driven current profile.

Research Activities carried out in parallel with FWP milestones

Goals denoted as Research Activities (RA) are important NSTX-U research activities that are not FWP milestones and may include substantial NSTX-U collaborator contributions.

RA(19-1):  Expand disruption prediction and avoidance capability for tokamaks

Description: Predicting and avoiding damaging plasma disruptions in fusion-producing tokamaks with high reliability is the present “grand challenge” in tokamak stability research. Meeting this significant goal requires data from a variety of tokamaks and the use of various physical models, analyses, and control methods. The present milestone will greatly expand automated disruption event characterization and forecasting (DECAF) that identifies chains of events that lead to disruptions. Once these chains are determined, methods of breaking the chains using all available control actuators can be defined. Key milestone deliverables include the expansion of the DECAF capability to accurately define disruption event chains across multiple devices and quantitatively increasing the reliability in predicting disruptions with low false positive rate. Through arranged collaborations, the analysis will include input from both national and international tokamaks (e.g. data from DIII-D, KSTAR, MAST-U, NSTX/NSTX-U), which is critical to produce reliable DECAF analysis applicable to ITER and future devices. DECAF code analysis has successfully demonstrated predictive models for global MHD mode onset and automatic determination of rotating tearing mode activity, rotation bifurcation events, and mode locking precursors to disruptions. These successes will be significantly expanded through the further development of physics modules and machine learning capabilities that address the dominant causes of disruptions across several experimental tokamak databases as stated above. Such development will produce and leverage new analysis capabilities and reduced models of results computed by stability analysis codes such as resistive and ideal DCON, M3D-C1, kinetic MHD analysis codes (e.g. MISK), and expanded kinetic equilibrium reconstruction capability. For example, validated physics models and analysis techniques determining tearing mode stability and island growth or decay would be coupled to models of torque balance and applicable code analysis will be tested for their predictive capability of rotating MHD mode locking. A range of models and analyses evaluating density limit-induced disruptions (both low and high) will be evaluated. Technical causes of disruptions will also be assessed. In support of the NSTX-U recovery effort and device restart, these results will be applied to identify how device actuators can be best used to avoid disruptions (e.g. restarting proportional gain and model-based, active n = 1 mode control (with synthetic diagnostics) which will provide important suppression of the device error field, its amplification, and mode control; starting rotation profile control for instability avoidance), and to improve the prediction of disruptions in real-time to inform and directly aid significant improvement of the NSTX-U controlled plasma shutdown handler capability.

RA(19-2):  Assess impact and importance of H species in HHFW-heated NSTX-U full-field plasmas

Description: The goal of NSTX-U is to operate at full field (B = 1 T) for 5 seconds. For this magnetic field, the first and second harmonics of hydrogen (H) are located at the high-field side and in the core plasma, respectively. In principle, part of the high-harmonic fast-wave (HHFW) injected power can be absorbed by the H population reducing the electron and/or the fast-ion heating. For this reason, full wave simulations of NSTX-U scenarios with different H concentrations will be performed. Plasma scenarios with and without neutral beam injection (NBI) will be considered. Furthermore, the possible impact of the tail in the H distribution function (at the 2nd ion cyclotron harmonic) to the electron heating will be investigated by using the combination of RF full wave and Fokker-Planck codes. Finally, an investigation of NSTX-U scenarios in which HHFW might modify either the electron or the ion temperature (through H species) will be analyzed, as it is of particular interest also for transport studies.