JRT-2018: Conduct research to test predictive models of fast ion transport by multiple Alfvén eigenmodes. (Led by NSTX-U)
Fusion alphas and injection of energetic neutral particle beams provide an important source of heating and current drive in advanced tokamak operating scenarios and burning plasma regimes. Alfven eigenmode (AE) instabilities can cause the redistribution or loss of fast ions and driven currents, as well as potentially decreasing fusion performance and leading to localized losses. Measured fast ion fluxes in DIII-D and NSTX-U plasmas with different levels of AE activity will be used to determine the threshold for significant fast ion transport, assess mechanisms and models for such transport, and quantify the impact on beam power deposition and current drive. Measurements will be compared with theoretical predictions, including quantitative fluctuation data and fast ion density, in order to validate models and improve understanding of underlying mechanisms. Model predictions will guide the development of attractive operating regimes.
R(18-1): Modeling of high divertor heat flux mitigation in NSTX-U
Description: By operating at twice the plasma current, toroidal field, and heating power relative to NSTX (with IP ≤ 2 MA, PNBI ≤ 10 MW, Bt ≤ 1 T), NSTX-U will access narrow SOL widths and high parallel heat fluxes in H-mode discharges. To guide heat flux mitigation development, modeling will be performed using multi-fluid two-dimensional transport codes, such as SOLPS and UEDGE, to study various techniques, individually and in combinations, including (1) magnetic balance (i.e. double-null operation), (2) radiative (partially detached) divertor operation, and (3) the high-flux expansion divertor. Detachment operational space for the highest SOL power and IP will be studied as a function of density and divertor seeding species, e.g. D2, CD4, Ne, Ar, and Li, in the standard and high flux expansion configurations. Detailed charge state resolved impurity models and realistic boundary conditions will be used in these simulations. Further, the impact of classical electromagnetic drifts on particle and radiated power distributions will be studied. Resources and time permitting, the interplay between density control via divertor cryo-pumping and detachment control will also be explored. The obtained results should enable further development of high performance spherical tokamak plasma scenarios with acceptable divertor power exhaust solutions and aid in the validation of divertor power and particle exhaust models for ITER and FNSF.
R(18-2): Performance evaluation of high-Z candidate design and materials for NSTX-U
Description: The NSTX-U has a multi-year program to transition the interior plasma-facing components from the existing graphite materials to metallic substrates. This transition is expected to mimic conditions expected in future power reactors as well as enable more reactor-relevant studies of lithium as a plasma-facing component. The NSTX-U conducted a project in FY15-16 to develop a continuous row of high-Z tiles for installation in the NSTX-U outboard divertor. The design was optimized for high-heat flux performance under the criteria that the existing mounting scheme be re-used. The analysis conducted during the project highlighted certain critical thermo-mechanical properties as well as the need for cyclic fatigue data in stress and temperature conditions expected in the NSTX-U. This research milestone will undertake accelerated-time testing of these relevant properties in a range of experimental devices. In particular, the impact of surface recrystallization and stress concentrators will be evaluated to determine if there are significant impacts on the overall life-time of refractory-metal components. Prototype diagnostics relevant to the high-Z material program will also be tested in order to further optimize the NSTX-U implementation. The data obtained in this research milestone will be used to validate the engineering efforts conducted in the high-Z design project and provide additional guidance for future, high-heat flux upgrades of the NSTX-U plasma-facing components. Models of boron- and lithium-coated metals will be further developed and compared to experimental results as appropriate.
R(18-3): Develop Tearing Mode Stability Analysis and Avoidance Strategies for NSTX-U
Description: Tearing modes (TMs) represent one of the most common causes of disruptions in high-performance tokamak discharges. Therefore, avoiding or mitigating TMs is a critical issue for progress toward steady-state, high-beta operation in both present and future devices. The goal of this milestone is to develop new techniques and refine existing ones for tearing stabilization and control on NSTX-U by developing improved physics understanding of the factors influencing tearing mode stability. These factors include aspect ratio, plasma shaping, alteration of the current, safety factor, rotation, and pressure profiles, and the application of non-axisymmetric fields. Existing experimental data from NSTX successfully utilized high plasma elongation (up to 3) and plasma profile modification due to lithium wall conditioning to successfully stabilize tearing modes. These results, new data from the FY-16 run campaign of NSTX-U, and data obtained via collaboration on other facilities will be analyzed to identify trends in the stability of tearing modes as a function of the controllable parameters stated above. Analysis using resistive and extended-MHD codes will be compared to experimental results in order to assess the capability to predict and model the empirically observed stability dependencies, and to provide physical insight into the physics of tearing mode stabilization in spherical torus geometry. Improved physics understanding, constrained by control and measurement capabilities, will be used to enable new techniques for tearing stabilization in NSTX-U.
R(18-4): Optimization of the energetic particle distribution function for improved plasma performance
Description: The improved neutral beam injection (NBI) capabilities that are available on NSTX-U enable a more flexible tailoring of the fast ion distribution function resulting from NBI. This milestone will explore the use of different NBI sources and timing of NB injection schemes to improve plasma performance and reproducibility by affecting fast ion-driven instabilities, e.g. through their mitigation or suppression. A main focus of this study is the early phase of the discharge (current ramp-up and early flat-top), during which strong fast ion-driven activity is destabilized as observed in most NSTX-U shots from the FY-16 experimental campaign. Instabilities include toroidal and reversed-shear Alfvénic modes (TAE/RSAE) as well as energetic particle modes and fishbones. The effect of sawteeth on the fast ion distribution function and NB current drive during the stationary phase of L-mode NSTX-U discharges will also be examined. These instabilities have the potential to cause substantial fast ion redistribution, thus affecting the overall efficiency of NB heating and current drive. Thus, if not properly accounted for in simulation codes, the effects of fast ion driven instabilities make the discharge evolution difficult to predict. Work within the Energetic Particle TSG will leverage and contribute to scenario development activities by the Advanced Scenarios and Control TSG. Once a suitable ramp-up scenario is identified, AE stability will be assessed for typical NSTX-U ramp-up scenarios. The analysis will include exploration of different NBI combinations and timing in time-dependent simulations to identify the optimum NB mix and resulting safety factor and current profiles that lead to reduced mode activity. Scenario development will rely on the TRANSP code. TRANSP analysis will be assisted by results from the NOVA/NOVA-K and ORBIT codes and from reduced models such as the ‘kick’ and Resonance-broadening Quasi-linear (RBQ) models to infer the mode stability.
R(18-5): Assess importance of multi-scale effects in NSTX/NSTX-U turbulent transport
Description: Electron scale (kqri >> 1) ETG turbulence has been predicted to be important in various NSTX and NSTX-U L-mode and H-mode plasmas. In some of these cases, gyrokinetic simulations indicate that ion scale (kqri < 1) turbulence should be completely suppressed by E´B shear such that only electron scale turbulence should remain. However, in many cases the ion scale turbulence is not predicted to be completely suppressed, or the predicted transport from ETG is insufficient to account for experimental electron heat flux levels. This milestone will focus on using gyrokinetic simulations and recently updated reduced models to clarify when and where ETG can be an important contributor in ST transport, and to begin addressing whether ETG can be considered in isolation from ion scale turbulence. Single-scale nonlinear gyrokinetic simulations will be used to predict transport based on experimental profiles, as done previously. Systematic parameter scans in driving gradients and E´B shearing rate will then be used to better quantify when ion scale transport dominates electron-scale transport, or vice versa. The region between these limits of single-scale transport should help demarcate where multiscale effects are expected to be important. To help guide the more computationally expensive gyrokinetic simulations, the TGLF reduced transport model will be used to predict similar boundaries in gradient-E´B shearing rate parameter space. Although the TGLF model has not yet been validated for ST parameters, it has been recently updated to recover the importance of multi-scale effects in conventional tokamak plasmas. In combination, we expect the results of the TGLF modeling and single-scale gyrokinetic simulations will be used to identify ST cases most suitable for multi-scale simulations. Initial results from simulations that begin to approach the multi-scale limit will be pursued to identify any computational challenges that may require special consideration.
R(18-6): Simulation framework development for NSTX-U high-performance scenarios
Description: Building upon the FY-2017 milestone for analysis and modelling of current ramp-up dynamics, a simulation framework will be developed that links the plasma control algorithms with a time-dependent plasma model capable of evolving the plasma boundary, current profile, and stored energy based on reduced models. This simulation framework will be used to develop and optimize real-time control tools and provide guidance on target scenarios that offer a robust discharge evolution that enables high Ip operation. During FY2017, initial tests of this framework will have been used to develop and test shape control algorithms using the discharge evolution realized on NSTX-U. Supporting analysis of NSTX-U data in FY17 will aid in the development of reduced models that would be used as a basis for predictive simulations for ramp-up scenarios. The simulation framework will include dynamic reduced models capable of driving predictive solutions. The reduced models will be compared against more comprehensive predictive plasma transport models where the free-parameters of the model are constrained by benchmarking with experimental results. MHD and fast-ion stability of achieved and proposed scenarios will be evaluated in order to identify constraints on the heating and q-profile during the ramp-up phase that would improve global stability. During FY2017 and extending into this FY2018, collaboration with MAST-U will be fostered to develop and apply simulation tools in order to accelerate the progress in establishing the physics basis for steady-state, high-beta ST devices.
R(18-7): Assess Transient CHI Current Start-up Potential for NSTX-U
Description: Solenoid-free plasma initiation would likely be required in a low aspect ratio FNSF/CTF facility to enable operation with minimal inboard neutron shielding. Transient Coaxial Helicity Injection (CHI) is a leading candidate method for solenoid-free plasma initiation in the ST configuration and has already demonstrated 150-200kA of closed-flux plasma current generation in NSTX. The present understanding of the current scaling for transient CHI suggests that the current generation potential is directly proportional to the device injector flux capability. The injector flux capability on NSTX-U is quite substantial. However, injecting more flux requires higher injector current capability and higher voltage, while simultaneously minimizing the influx of impurities at the higher injector parameters. The scaling of the injector flux to both the current generation and the enclosed poloidal flux is needed for projecting to next-step device designs, such as the ST-FNSF. As the amount of injected poloidal flux in the CHI discharge increases, the inductive drive from the decaying CHI plasma should increase, and allow the CHI plasma intrinsic electron temperature to increase. Simulations using the TSC and NIMROD codes will be performed to assess the maximum injector flux that could be used on NSTX-U, the resulting increases to the electron temperature, and the dependence of 3-D reconnection processes on injector parameters as the amount of injector flux is increased to the limits possible on NSTX-U and ST-FNSF. Supporting experiments will be conducted on the QUEST ST in Japan to establish transient CHI capability in an alternate complementary electrode configuration, using metallic electrodes. An initial test of electron heating due the application of high-power electron cyclotron heating in QUEST will also be conducted if technically ready.
R(18-8): Toward self-consistent calculations of fast wave and energetic-ion interactions
Description: Self-consistent modeling of the interactions between fast waves and fast ions, introduced either from neutral beam injection (NBI) or from fusion-generated alpha particles, is important for both present-day experiments and also for ITER. The fast-ion population changes the wave propagation and absorption, while the wave damping on fast-ions modifies their distribution. The latter implies that fast-wave heating could impact and perhaps give leverage over Alfvénic activity. Specific to NSTX-U, simultaneous high-harmonic fast-wave (HHFW) heating and NBI is desirable for experiments in turbulence, impurity transport, and Alfvénic activity. However, because of the lower toroidal field of the spherical tokamak, fast-wave heating may accelerate fast ions to loss orbits, and this power-loss mechanism must be studied and then minimized. To this end, self-consistent calculations of the wave fields and the fast-ion distribution function will be pursued by (1) upgrading a full-wave solver to compute the wave fields for arbitrary ion distributions, and (2) iteratively evaluating the full-wave solver with the Monte-Carlo particle code NUBEAM. A recent extension of the full-wave code TORIC v.5 now computes non-Maxwellian ion effects. The TORIC extension will be verified for the standard cases and used to explore effects of independently varying the parallel and perpendicular temperatures for both the ion-cyclotron minority and HHFW regimes. Then, the TORIC extension will be applied to NSTX discharges using the NSTX distribution function obtained from NUBEAM, giving more self-consistent and accurate calculations of the HHFW power deposition profile as well as the impact of HHFW heating the fast ion distribution function. Attention will be paid to possible fast-wave interactions with the energetic-particle-driven instabilities, as observed in previous NSTX experiments. The coupling between TORIC v.5 and NUBEAM will be implemented in the TRANSP framework and could be used for other fusion experiments.