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FY2020 Research Milestones

JRT20:    Validation of neoclassical and turbulent impurity transport models in reactor relevant plasma conditions

Accumulation of impurities, ranging from light ions (helium ash) to high-Z (such as tungsten) can adversely impact the reactivity of the fusion core through fuel dilution and excessive radiation.  To inform operation of ITER and beyond, transport of impurities from the divertor to the core will be studied, particularly as parameters that are expected to impact the relative balance of turbulent versus neoclassical impurity transport are varied. Experiments will introduce a wide range of low to high Z impurities, while turbulence and transport properties are documented. Integrated modeling tools will be used to validate theoretical models and interpret the physical mechanisms of transport in the core, divertor, and scape-off layer.

R(20-1):  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 (Bt~ 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 FY2020, 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 FY2020, emphasis will be placed on collaboration on MAST-U to support the core transport and pedestal structure research goals of this milestone.


R(20-2): Commission operational tools that enable high-performance discharges in NSTX-U

Description:  NSTX-U is designed to develop the physics and technical basis required for stationary, long pulse, high non-inductive fraction operation in a low-aspect-ratio tokamak. A major research goal during the FY20 commissioning campaign on NSTX-U is to develop high-performance H-mode scenarios that simultaneously exceed the current (Ip > 1.4 MA) and magnetic field strength (BT > 0.55) achieved on NSTX. A critical component of these scenarios is maximizing the achievable elongation at low internal inductance (li). This effort will build on the ramp-up simulation development completed in FY19 that identifies ramp-up scenarios that optimize the achievable elongation.  Dedicated experiments will quantify the vertical and MHD stability limits in the ramp-up phase in order to compare to the simulation results and identify avenues for potential expansion of these limits through new scenario or control tools. Building upon the results of the FY16 NSTX-U run campaign and the FY17-18 milestones on error field identification and correction, a re-assessment of low-n error fields, mode-locking, and optimal error field correction will be made.  Further, RWM control and dynamic error field correction algorithms using both proportional and state-space n ≥ 1 feedback schemes will be implemented taking advantage of the spectrum flexibility provided by the 2nd SPA power supply. This effort will enable access to large bN/li, which is critical for high-current H-mode scenarios. Resonant field amplification measurements, ideal MHD stability codes, and kinetic stability analysis will be used to evaluate the no-wall and disruptive stability limits. These physics and operational tools will be combined to enable new plasma operating scenarios and to make an initial assessment of the non-inductive current drive fraction across a range of toroidal field, plasma density, boundary shaping, and neutral beam parameters. During FY20, significant effort will be put toward commissioning real-time diagnostics and scenario-control-relevant actuators for these experiments on NSTX-U. This may include the integration of ELM-pacing actuators in order to control the accumulation of impurities.  If NSTX-U cannot support plasma operations during FY20, additional emphasis will be placed on collaboration on MAST-U to support the high-current access, shape control, error-field correction, and stability analysis research goals of this milestone.


R(20-3):  Integrated Disruption Modeling for NSTX-U

Description (to be fleshed out): In order to better understand electromechanical stresses, thermal loads, and runaway electron generation during disruptions, nonlinear simulations of both the thermal quench and current quench phases of disruptions and vertical displacement events (VDE) will be carried out for NSTX and NSTX-U discharges. These 3D MHD simulations will be performed using an integrated physics model that includes models for impurity radiation and transport, halo and eddy currents, and runaway electron generation. Non-axisymmetric instabilities and magnetic stochasticity will also be included. The effects of mitigation using massive gas injection or pellet injection will be considered.  Simulations will be validated by comparing results with experimental measurements of vertical displacement, magnetic probes, shunt tile currents, and the change in stored thermal energy.

R(20-4): 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 FY20, 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 sufficient plasma operations during FY2020, 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(20-1):  Multi-modes correction of error fields in NSTX-U and tokamaks

Description: A small non-axisymmetric error field (EF) can cause a disruptive field penetration or mode locking instability and therefore must be properly controlled to achieve high performance in tokamaks. Although the empirical scaling based on a single dominant n=1 mode approximation has been partially successful, multi-mode EF, including n > 1 modes, can also degrade plasma confinement depending on the operating regime, as has recently been demonstrated in various tokamaks. For example, EF identification and correction in the 2016 NSTX-U campaign showed that the plasma response to the inner-TF and PF5 n = 1 misalignment can change significantly as the current profile evolves, and therefore these EFs are not easily correctable by standard control algorithms.  This sensitivity of the plasma response is due to the coupling to the m = 1 component, the generically complex phase variation of the high-field-side (HFS) EFs, and also possibly the shift of the magnetic axis. These NSTX-U EF data will be further investigated and compared with other new experiments from COMPASS, DIII-D, MAST-U, KSTAR, and EAST, in order to develop single-mode and multi-mode EF threshold scaling and correction strategies for future tokamak devices. The regime dependence of EF thresholds for error field penetration and mode locking will be studied by applying leading theories relevant for each linear or non-linear, and two-fluid or drift-kinetic MHD, and also by applying hybrid and extended MHD codes to study dynamics and bifurcations of islands due to EFs predicted in theory. Improved EF correction criteria will be developed for use in ITER, MAST-U, and NSTX-U. Coil metrology, EF characterization, and non-axisymmetric response experiments will be revised for newly assembled NSTX-U, and multi-n correction strategies against penetration and TM/NTM mode locking will be developed and tested.


RA(20-2):  Application of expanded disruption prediction and avoidance capability for NSTX-U

Description: Disruption event characterization and forecasting (DECAF) capabilities will be expanded under NSTX-U research activity RA(19-1). These new capabilities will be interfaced with the NSTX-U plasma control system (PCS) along with global mode avoidance tools for disruption avoidance. In support of the NSTX-U device restart, disruption forecasting results will be applied to identify how device actuators can be best used to avoid disruptions. Understanding from DECAF analysis will be implemented into the NSTX-U off-normal event shutdown handler to allow controlled plasma shutdown as desired. Proportional gain and model-based, active n = 1 mode control (with synthetic diagnostics) will be prepared and expanded. These active control capabilities for NSTX-U will provide important suppression of the device error field and its amplification, and mode control. To avoid resistive wall mode onset, rotation profile control algorithm implementation will be started to enable disruption avoidance. The RWM state-space controller observer will also be tested as a criterion for the NSTX-U shutdown handler capability. Real-time MHD spectroscopy (successfully applied to detect the plasma stability and multi-mode plasma response in DIII-D and EAST experiments) will be investigated for use in NSTX-U to actively determine stability to MHD modes while the plasma is stable. The method will be tested and optimized for NSTX-U under this research activity further improving its efficacy for real-time use. The real-time output of the system will also be implemented as a component input to the NSTX-U shutdown handler.