FY2018 Research Milestones

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): Develop and benchmark reduced heat flux and thermo-mechanical models for PFC monitoring

Description: The NSTX-U Recovery Project will deploy new plasma facing components (PFCs) to meet updated heat exhaust requirements driven by narrower scrape-off-layer widths, increased heating power, and longer pulse durations relative to NSTX. Inter-shot monitoring or intra-shot control of heat flux to PFCs is anticipated for a range NSTX-U operating space, necessitating reduced models that can be run between shots or even in real-time. Monitoring requires a reliable instrumentation suite which can support or contradict model predictions and confirm PFC integrity. The goals of this milestone are three-fold: (1) Develop tools for pre-shot planning and confirmation of post-shot PFC thermal observations which use reduced models to predict time-evolving heat fluxes to shaped PFCs and estimate distances from engineering limits. Assess additional effort needed for implementation of reduced models in PCS. (2) Where feasible, benchmark reduced models against boundary physics (e.g. SOLPS, UEDGE) and finite element analysis (e.g. ANSYS) tools, and validate using experimental data from relevant tokamaks and results from Facility Milestone F(18-1). (3) Evaluate examples of discrete monitoring systems that are sufficient to capture the evolution of the PFCs relative to engineering limits. Compare the ability for different techniques (e.g. thermocouples vs. imaging) and technologies (e.g. near vs. long-wave infrared cameras) to achieve NSTX-U PFC monitoring objectives.

R(18-2): Develop simulation framework for spherical tokamak breakdown and current ramp-up

Description: Access to high-performance discharges in spherical tokamaks (STs) is sensitive to the first phase of the plasma discharge, including the initial generation of the plasma discharge (i.e. breakdown) and the increase of the plasma current, temperature and density to the target value (i.e. ramp-up). An optimized ramp-up scenario would minimize the internal inductance (l_i) and flux consumption while exhibiting reasonable resilience to expected variations in the experimental conditions. This milestone aims to develop computational tools that enable the optimization of the breakdown and ramp-up phase on NSTX-U and MAST-U. A reduced model computational framework, such as TOKSYS, will be developed in order to connect the real-time plasma control algorithms with a time-dependent model including the power supply capabilities, toroidal currents induced in the vessel structures and a free-boundary plasma equilibrium. This model will be used to gain insight into the vertical stability limits during ramp-up and the impact of power supply, wall structures and plasma parameters on the maximum stable elongation in STs. This framework will also be used to develop, test and optimize the real-time shape control algorithms in the ramp-up phase. A second framework approach will use comprehensive transport simulation, such as TRANSP, to investigate the evolution of the kinetic profiles during ramp-up as a function of the free parameters. The initial goal is to optimize the outer gap, density and neutral beam heating in the L- and H-mode phases of the ramp-up that minimize l_i and flux consumption while remaining within MHD and fast-ion stability limits. The results from the first reduced model framework will provide guidance on the target high-elongation shapes, while the second transport model framework will provide guidance on the evolution of equilibrium parameters such as l_i and β_N. This milestone also aims to develop a simulation framework for optimizing aspects of the inductive plasma breakdown, such as the null formation and the initial increase of the vertical fields, over the range of expected ohmic solenoid pre-charge and toroidal field currents. The proposed development of simulation frameworks described in this milestone will reduce the experimental time required to achieve a suitable optimization of the breakdown and ramp-up scenarios for accessing high-performance scenarios on NSTX-U and MAST-U.

R(18-3): Validate and further develop reduced transport models for electron thermal transport in ST plasmas

Description: The design of next generation spherical tori (STs) will be influenced by the scaling of energy confinement. While ion thermal transport is often near neoclassical levels in H-modes in ST plasmas, gyro-kinetic simulations have indicated a number of potential drift wave turbulence mechanisms that can influence electron thermal transport. Reduced transport models that capture the key physics and scaling of the computationally expensive first-principles gyro-kinetic simulations are required to more thoroughly validate the modeling against experimental data, which can then be used to infer the key physics that determines the overall energy confinement. A variety of reduced transport models based on drift wave turbulence have been developed and tested extensively for conventional tokamaks. These models encompass much of the physics expected to be important in STs, although they have been tested much less rigorously for ST parameters (low aspect ratio, high beta, strong flow). In order to improve the fidelity of reduced transport models (like TGLF, RLW and MMM), experimental NSTX, MAST and NSTX-U data will be used to examine predictions based on these models to assess their suitability for ST plasma. The physics accuracy of these fluid-based models will be also be qualified by comparing directly to first-principles gyro-kinetic simulations over a range of conditions. The dependence of electrostatic ITG and TEM instabilities on aspect ratio will be evaluated by comparing L-mode cases to established conventional aspect ratio conditions. Validation with high beta H-mode data will push the limits of the available reduced models to recover electromagnetic instabilities like MTM and KBM. A key outcome of this milestone will be to determine the ST physics regimes in which further model development is required. The first-principles gyro-kinetic simulations based on ST parameters will form the basis for enhancements of the TGLF reduced model.

R(18-4): Optimize energetic particle distribution function for improved plasma performance

Description: The improved neutral beam injection (NBI) capabilities that are available on NSTX-U enable a flexible tailoring of the fast ion distribution function resulting from NBI. In collaboration with DIII-D and MAST-U, this milestone will explore the use of different NBI sources and timing of NB injection 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 current ramp-up/early flat-top phase, during which strong fast ion-driven activity can be destabilized (cf. 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. Sawteeth during the stationary phase of L-mode NSTX-U discharges will also be included. All these instabilities have the potential to cause substantial fast ion redistribution, thus affecting the overall efficiency of NB heating and current drive. If not properly accounted for in simulations, 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, including the planned collaboration with MAST-U in FY17-18. Once a suitable target scenario is identified, AE and fishbone stability will be assessed. The analysis will include exploration of different NBI combinations (e.g. on- vs. off-axis) 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. Validation of the ‘kick model’ for scenarios with unstable fishbones will be conducted in collaboration with MAST-U. In collaboration with DIII-D, a recently developed criterion to predict the nonlinear behavior of Alfvénic instabilities (e.g. quasi-stationary vs. bursting/chirping) will be validated to gain further confidence in predictions of the fast ion transport instabilities can cause. Test particle simulations of fast ion scattering by plasma turbulence will be performed using the GTS code to assist the validation of the theoretical criterion for instability chirping.

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(18-1): Validation of non-axisymmetric plasma response modeling to address compatibility with core and edge constraints

Description: The application of non-axisymmetric magnetic fields to tokamaks is a useful technique for controlling plasma rotation, transport, and stability. Reliance on this technique in NSTX-U, ITER and future reactors requires sufficient understanding to design an applied field spectrum that will perform the desired actuation without inducing deleterious core MHD modes or displacing the heat flux footprint in high-performance discharges. In particular, non-axisymmetric fields may strongly affect the magnetic geometry in the divertor region, which may have important implications for closed-divertor configurations such as the Small-Angle Slot (SAS) on DIII-D and the Super-X divertor on MAST-U, especially regarding access to detached operation and the deposition of heat flux outside the divertor. Pitch-angle changes caused by non-axisymmetric fields may also have implications for the design of divertor tiles and acceptable pulse lengths/parameters in NSTX-U and other devices. Modeling of the plasma response to applied non-axisymmetric fields will be carried out to investigate these issues in MAST-U, DIII-D, and NSTX(-U) using ideal, resistive, and two-fluid models. The effect of non-axisymmetric fields on the divertor leg geometry, strike point splitting, and pitch angle at the plasma facing components will be calculated and provided as input for calculations of the heat and particle flux. This will go beyond work done for a previous milestone R(17-2) by considering a broader range of conditions, devices, and spectra of applied fields, to extend the scope of validation and provide input for potential future NSTX-U upgrades (such as NCCs). A range of n = 1 and n > 1 fields will be considered, consistent with the realizable coil configurations and intrinsic error fields on each device. The effect of profile changes due to non-axisymmetric fields on global MHD stability will also be considered, including non-ideal and kinetic effects, with a focus on high-performance scenarios. Modeling results will be validated against experimental data by comparing with measurements of the plasma response, divertor heat deposition, and observations of instabilities. These results will be used to inform the NSTX-U recovery process and operational constraints.

RA(18-2): Develop 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 and Fokker-Planck code CQL3D. 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 JET (including D-T scenarios), NSTX and ITER scenarios using the fast ion distribution function obtained from NUBEAM and/or CQL3D, giving more self-consistent and accurate calculations of the RF power deposition profile as well as the impact of RF heating on the fast ion distribution function. Attention will be paid to possible fast-wave interactions with the energetic-particle-driven instabilities in previous NSTX experiments and ITER scenarios, at half-field in particular, where the NBI species have an ion cyclotron resonance in the plasma. The coupling between TORIC v.5 and NUBEAM will be implemented in the TRANSP framework and could be used for other fusion experiments (such as ASDEX, EAST, KSTAR, etc.).

RA(18-3): Assess transient CHI current start-up potential for ST current initiation

Description: Solenoid-free plasma initiation is likely required in future ST devices such as a FNSF/CTF facility to enable operation at low aspect ratio 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 100-200kA of closed-flux plasma current generation in present STs. The present understanding of the current scaling for transient CHI suggests that the current generation potential is directly proportional to the device’s injector flux capability. At present there does not seem to exist a physics limit to the amount of poloidal-injected flux that could be used for start-up. Simulations using the TSC and NIMROD codes will be performed to assess the maximum injector flux that could be used in a ST configuration. The resulting increases to the electron temperature, and the dependence of 3-D reconnection processes on injector parameters as the amount of injector flux and the toroidal field is increased to levels capable of generating multi-MA level start-up currents will be examined. In particular, the 3-D MHD physics modeling is critical for investigating whether all of the needed current required for plasma sustainment could be generated by transient CHI alone. Using 3-D simulations, the feasible maximum injector flux will be determined. Another aspect that may be important for the implementation of CHI in a FNSF is the use simpler electrode structures, such as for example biasing an electrode with respect to a grounded vessel component. Scoping design studies for a ST-FNSF suggests that with the use of biased electrodes, the insulator could be easily protected from neutrons and that the insulator would maintain its integrity for the lifetime of the reactor. To test if such an electrode configuration is compatible with transient CHI operations, supporting experiments will be conducted on the QUEST ST in Japan to establish transient CHI capability in this, recently installed, alternate electrode configuration, using metal electrodes. An initial test of electron heating due the application of high-power electron cyclotron heating in QUEST may also be conducted. The NSTX-U device is also capable of supporting such an electrode configuration, for example by placing an insulated electrode below the inner upper divertor plate. Scoping studies would be conducted to investigate the design of such an alternate electrode configuration for NSTX-U. A design based on biased electrodes will be developed for a CHI system for PEGASUS that is capable of generating 200 – 300kA start-up currents, to the maximum levels that could be supported by the existing PF coils on PEGASUS. A feasibility study for the installation of a point source helicity injection on NSTX-U will also be conducted.