Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Iterative”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 307 records · Page 17

Accelerating and stabilizing the convergence of vibrational self-consistent field calculations via the direct inversion of the iterative subspace (vDIIS) algorithm

The vibrational self-consistent field (VSCF) method yields anharmonic states and spectra for molecular vibrations, and it serves as the starting point for more sophisticated correlated-vibration methods. Convergence of the iterative, non-linear optimization in VSCF calculations can be erratic or altogether unsuccessful, particularly for chemical systems involving low-frequency motions. In this work, a vibrational formulation of the Direct Inversion of the Iterative Subspace method of Pulay is presented and investigated. This formulation accounts for distinct attributes of the vibrational and electronic cases, including the expansion of each single-mode vibrational wavefunction in its own basis set. The resulting Direct Inversion of the Iterative Subspace method is shown to substantially accelerate VSCF convergence in all convergent cases as well as rectify many cases where Roothaan-based methods fail. Performance across systems ranging from small, rigid molecules to weakly bound molecular clusters is investigated in this analysis.

Chemistry↗

Benchmark between antenna code TOPICA, RAPLICASOL and Petra-M for the ICRH ITER antenna

ITER will be equipped with three plasma heating systems: neutral beam (NB), electron cyclotron (EC), and ion cy-clotron resonance heating (ICRH). The latter consists of two identical ICRH antennas to deliver 20 MW to the plasma (baseline, upgradable to 40 MW). ICRH will play a crucial role in the ignition and sustainment of burning plasmas in ITER. A high fidelity and robust modeling effort to understand the interaction of the IC waves with the scrape-off-layer (SOL) plasma is a very important aspect. Among the main important research topics, we have the assessment of the antenna loading for different plasma scenarios, the role of the lower hybrid resonance in front of the antenna and how to include it in our models, and the RF sheath boundary conditions to evaluate the antenna impurity generation. In this work, we tackle the first of these by reporting on ICRF simulations employing the Petra-M code, which is an electromagnetic simulation tool for modeling RF wave propagation based on MFEM [http://mfem.org] for the ITER ICRH antenna. Moreover, a benchmark between the well tested antenna codes TOPICA, RAPLI-CASOL, which is based on COMSOL [www.comsol.com], and the Petra-M code is also presented. In conclusion, S- and Z-matrices and wave electric field are compared showing an excellent agreement among these codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

The motional Stark effect diagnostic for ITER

An overview of the plans for the motional Stark effect (MSE) diagnostic installation on the International Thermonuclear Experimental Reactor (ITER) is presented. The MSE diagnostic uniquely provides spatially localized magnetic field measurements inside the plasma. These are used to constrain equilibrium reconstructions to determine q(r), the safety factor as a function of minor radius. Meeting the system requirements to deliver q-profiles and related quantities with the specified radial resolution of 20 points over the minor radius, 10 ms time resolution, and better than 10% accuracy is challenging. MSE systems observe the D/H-α emission near 656.3 nm from neutral beams. As the beam atoms traverse the magnetic field, B⃗, at high velocity, v⃗, they experience a Lorentz electric field, v⃗×B⃗, which causes the spectral emission to be split and polarized due to the Stark effect. Traditional MSE-LP (line polarization) measurements determine the direction of the magnetic field in the observation volume using polarimetric analysis of the detected light. The harsh conditions of ITER are expected to deposit thin films of contaminants on the first mirror, which would alter the polarization state of reflected light significantly. On ITER, the combination of high magnetic field strength and high energy beams makes the Stark spectrum resolution suitable for the determination of the magnetic field magnitude from the line shift, so this approach has been selected. Every aspect of the measurement system must be planned for the burning plasma environment and carefully analyzed ahead of time. Current status and plans for the system are presented.

Instruments & Instrumentation↗

A Weighted Least-Squares Transport Equation Compatible with Source Iteration and Voids

We present that second-order forms of the transport equation allow the use of continuous finite elements (CFEMs). This can be desired in multiphysics calculations where other physics require CFEM discretizations. Second-order transport operators are generally self-adjoint, yielding symmetric positive-definite (SPD) matrices, which allow the use of efficient linear algebra solvers with an enormous advantage in memory usage. Least-squares (LS) forms of the transport equation can circumvent the void problems of other second-order forms but are almost always nonconservative. Additionally, the standard LS form is not compatible with discrete ordinates method (SN) iterative solution techniques such as source iteration. A new form of the LS transport equation has recently been developed that is compatible with voids and standard SN iterative solution techniques. Performing nonlinear diffusion acceleration (NDA) using an independently differenced low-order equation enforces conservation for the whole system and makes this equation suitable for reactor physics calculations. In this context, “independent” means that both the transport and low-order solutions converge to the same scalar flux and current as the spatial mesh is refined, but for a given mesh, the solutions are not necessarily equal. In this paper we show that introducing a weight function into this LS equation improves issues with causality and can render our equation equal to the self-adjoint angular flux (SAAF) equation. Causality is a principle of the transport equation that states that information travels only downstream along characteristics. This principle can be violated numerically. We show how to limit the weight function in voids and demonstrate the effect of this limit on accuracy. Using the C5G7 benchmark, we compare our method to the SAAF formulation with a void treatment (SAAFτ) that is not self-adjoint and has a nonsymmetric coefficient matrix. Finallyl, we show that the weighted LS equation with NDA gives acceptable accuracy relative to the SAAFτ equation while maintaining a SPD system matrix.

97 MATHEMATICS AND COMPUTING↗

ITER ECH Switch Design and Analysis

The electron cyclotron heating system (ECH) on ITER uses 24 evacuated microwave transmission lines carrying up to 1.4 MW of power each at 170 GHz to provide resonance heating of electrons in the ITER plasma and to enable plasma current drive. A critically important component in this system is the microwave switch that allows the microwaves to be directed from the gyrotrons to either dummy loads or between launchers in the upper and equatorial ports of the ITER tokamak while maintaining the vacuum integrity of the transmission lines. A moveable, water-cooled CuCrZr mirror is used to redirect the microwave transmission between two orthogonal waveguides. In this paper, we describe the optimized design of the mirror cooling passages produced by computational fluid dynamics analysis using ANSYS CFX with k-ε and k-ω shear stress transport turbulence models, and verify that the design parameters for mass flow rate, inlet temperature and pressure are adequate for good thermomechanical performance. Non-uniform heating of the mirror face from the incident microwaves induces deflections that should be less than 25 microns to meet the integrated transmission line efficiency specification. In the current 1.4 MW switch design, 0.03 kg/s of 36°C water at 10 bar inlet pressure can remove the 2660 W of ohmic heating in the mirror produced by the elliptical polarization power and maintain the surface temperature below 150°C. The water delta-T is 21°C with a 0.5 bar pressure drop in the mirror. Finally, the maximum predicted displacement in the center of the mirror face is less than 25 μm.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Pressure Response Optimization of an Eddy Current-Driven Flyer Plate Valve for the ITER Shattered Pellet Injection System

One technique for mitigating disruptions in a tokamak is shattered pellet injection (SPI). SPI is a process in which a large solid pellet consisting of deuterium, neon, or argon is desublimated in a pipe gun barsrel and launched downstream. Pellets are shattered just before entering the plasma by an impact with an angled tube. Injection of these materials into the plasma radiates stored thermal energy, limits current decay rates, suppresses the generation of runaway electrons, and dissipates runaway electrons if necessary. A critical element of the SPI system is a fast-acting valve that releases high-pressure gas to dislodge and accelerate pellets directly, or indirectly via a mechanical punch. A prototype valve sized for the ITER SPI system has been designed and fabricated. A pulsed high-voltage power supply energizes the valve’s internal magnetic coil, which induces eddy currents in the adjacent flyer plate resulting in a repulsive force between the flyer plate and the coil. The flyer plate action lifts a valve seat, allowing high-pressure gas to flow from the valve plenum to the downstream (breech) location of the pellet or mechanical punch. All of the valve’s internal components are designed to operate in ITER-level static background magnetic fields. Here, a study was conducted to optimize the downstream pressure response for a range of valve sizes and operating pressures. In particular, the study analyzes the breech pressure response associated with varying plenum pressures as well as varying breech volumes. A computational fluid dynamics simulation was built in STAR-CCM+ and validated against data from laboratory experiments. The resulting simulation outputs, in the form of downstream responses for a variety of initial plenum pressures and breech volumes, will be used as a complement to experimental data to ensure the pressure pulse is suitable for pellet survivability. These data, combined with studies on pellet shear strength and shock response, will be applied to optimization of overall operating parameters of the ITER SPI system.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DIII-D research to address key challenges for ITER and fusion energy

DIII-D has made significant advances in the scientific basis for fusion energy. The physics mechanism of resonant magnetic perturbation (RMP) edge localized mode (ELM) suppression is revealed as field penetration at the pedestal top, and reduced coil set operation was demonstrated. Disruption runaway electrons were effectively quenched by shattered pellets; runaway dissipation is explained by pitch angle scattering. Modest thermal quench radiation asymmetries are well described NIMROD modeling. With good pedestal regulation and error field correction, low torque ITER baselines have been demonstrated and shown to be compatible with an ITER test blanket module simulator. However performance and long wavelength turbulence degrade as low rotation and electron heating are approached. The alternative QH mode scenario is shown to be compatible with high Greenwald density fraction, with an edge harmonic oscillation demonstrating good impurity flushing. Discharge optimization guided by the EPED model has discovered a new super H-mode with doubled pedestal height. Lithium injection also led to wider, higher pedestals. On the path to steady state, 1 MA has been sustained fully non inductively with β N = 4 and RMP ELM suppression, while a peaked current profile scenario provides attractive options for ITER and a β N = 5 future reactor. Energetic particle transport is found to exhibit a critical gradient behavior. Scenarios are shown to be compatible with radiative and snowflake diverter techniques. Physics studies reveal that the transition to H mode is locked in by a rise in ion diamagnetic flows. Intrinsic rotation in the plasma edge is demonstrated to arise from kinetic losses. New 3D magnetic sensors validate linear ideal MHD, but identify issues in nonlinear simulations. Detachment, characterized in 2D with sub-eV resolution, reveals a radiation shortfall in simulations. As a result, future facility development targets burning plasma physics with torque free electron heating, the path to steady state with increased off axis currents, and a new divertor solution for fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

High internal inductance for steady-state operation in ITER and a reactor

Increased confinement and ideal stability limits at relatively high values of the internal inductance (ℓ i ) have enabled an attractive scenario for steady-state tokamak operation to be demonstrated in DIII-D. We achieved normalized plasma pressure in the range appropriate for a reactor in high elongation and triangularity double-null divertor discharges with beta(N) approximate to 5 at ℓ i approximate to 1.3, near the ideal n = 1 kink stability limit calculated without the effect of a stabilizing vacuum vessel wall, with the ideal-wall limit still higher at β N > 5.5. Moreover, confinement is above the H-mode level with H 98(y,H- 2) approximate to 1.8. At q 95 approximate to 7.5, the current is overdriven, with bootstrap current fraction f BS approximate to 0.8, noninductive current fraction f NI > 1 and negative surface voltage. For ITER (which has a single-null divertor shape), operation at ℓ i approximate to 1 is a promising option with f BS approximate to 0.5 and the remaining current driven externally near the axis where the electron cyclotron current drive efficiency is high. This scenario has been tested in the ITER shape in DIII-D at q 95 = 4.8, so far reaching f NI = 0.7 and f BS = 0.4 at beta N approximate to 3.5 with performance appropriate for the ITER Q=5 mission, H 89 beta N /q 2 95 approximate to 0.3. Finally, modeling studies explored how increased current drive power for DIII-D could be applied to maintain a stationary, fully noninductive high ℓ i discharge. Stable solutions in the double-null shape are found without the vacuum vessel wall at beta N = 4, ℓ i = 1.07 and f BS = 0.5, and at beta N = 5 with the vacuum vessel wall.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Extending the physics basis of quiescent H-mode toward ITER relevant parameters

Recent experiments on DIII-D have addressed several long-standing issues needed to establish quiescent H-mode (QH-mode) as a viable operating scenario for ITER. In the past, QH-mode was associated with low density operation, but has now been extended to high normalized densities compatible with operation envisioned for ITER. Through the use of strong shaping, QH-mode plasmas have been maintained at high densities, both absolute ($\bar{n}$ e ≈ 7 × 10 19 m ₋3 ) and normalized Greenwald fraction ($\bar{n}$ e /n G > 0.7). In these plasmas, the pedestal can evolve to very high pressure and edge current as the density is increased. High density QH-mode operation with strong shaping has allowed access to a previously predicted regime of very high pedestal dubbed “Super H-mode”. Calculations of the pedestal height and width from the EPED model are quantitatively consistent with the experimentally observed density evolution. The confirmation of the shape dependence of the maximum density threshold for QH-mode helps validate the underlying theoretical model of peeling- ballooning modes for ELM stability. In general, QH-mode is found to achieve ELM- stable operation while maintaining adequate impurity exhaust, due to the enhanced impurity transport from an edge harmonic oscillation, thought to be a saturated kink- peeling mode driven by rotation shear. In addition, the impurity confinement time is not affected by rotation, even though the energy confinement time and measured E×B shear are observed to increase at low toroidal rotation. Together with demonstrations of high beta, high confinement and low q 95 for many energy confinement times, these results suggest QH-mode as a potentially attractive operating scenario for the ITER Q=10 mission.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

The ITER 3D Magnetic Diagnostic Response to Applied n=3 and n=4 RMP's

The ITER magnetic diagnostic response to applied n=3 and n=4 RMPs has been calculated for the 15MA scenario. The VMEC code was utilized to calculate free boundary 3D ideal MHD equilibria, where the non-stellarator symmetric terms were included in the calculation. This allows an assessment to be made of the possible boundary displacements due to RMP application in ITER. As the VMEC code assumes a continuous set of nested flux surface, the possibility of island and stochastic region formation is ignored. At the start of the current at-top (L-Mode) application of n = 4 RMP's indicates approximately 1 cm peak-to-peak displacements on the low field side of the plasma while later in the shot (H-mode) perturbations as large as 3 cm are present. Forward modeling of the ITER magnetic diagnostics indicates significant non-axisymmetric plasma response, exceeding 10% the axisymmetric signal in many of the flux loops. Magnetic field probes seem to indicate a greater robustness to 3D effects but still indicate large sensitivities to 3D effects in a number of sensors. Forward modeling of the diagnostics response to 3D equilibria allows assessment of diagnostics design and control scenarios.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Joint DIII-D/EAST research on the development of a high poloidal beta scenario for the steady state missions of ITER and CFETR

Experimental and modeling investigations on the DIII-D and EAST tokamaks show the attractive transport and stability properties of fully noninductive, high poloidal-beta (β P ) plasmas, and their suitability for steady-state operating scenarios in ITER and CFETR. A key feature of the high-β P regime is the large-radius (r>0.6) internal transport barrier (ITB), often observed in all channels (ne, Te, Ti, rotation), and responsible for both excellent energy confinement quality and excellent stability properties. Experiments on DIII-D have shown that, with a largeradius ITB, very high β N and β P values (both≥4) can be reached by taking advantage of the stabilizing effect of a nearby conducting wall. Synergistically, higher plasma pressure provides turbulence suppression by Shafranov shift, leading to ITB sustainment independent of the plasma rotation. Experiments on EAST have been used to assess the long pulse potential of the high-β P regime. Using RF-only heating and current drive, EAST achieved minute-long fully noninductive steady state H-mode operation with strike points on an ITER-like tungsten divertor. Improved confinement (relative to standard H-mode) and steady state ITB features are observed with a monotonic q-profile with q min ~1.5. Separately, experiments have shown that increasing the density in plasmas driven by lower hybrid wave broadens the q-profile, a technique that could enable a large radius ITB. These experimental results have been used to validate MHD, current drive, and turbulent transport models, and to project the high-β P regime to a burning plasma. These projections suggest the Shafranov shift alone will not suffice to provide improved confinement (over standard H-mode) without rotation and rotation shear. However, increasing the negative magnetic shear (higher q on axis) provides a similar turbulence suppression mechanism to Shafranov shift, and can help devices such as ITER and CFETR achieve their steady-state fusion goals.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Impact of window temperature variations on ITER toroidal interferometer and polarimeter (TIP) measurements

Abstract The toroidal interferometer and polarimeter (TIP) is one of the primary electron density diagnostics on ITER. To meet measurement requirements over several thousand seconds, environmental factors such as changes in the window temperature, air temperature, and humidity, which can cause uncompensated phase drifts, must be minimized. This paper reports measurements of the phase shift induced by ZnSe and BaF₂ windows as their temperature is varied utilizing the TIP prototype. To accomplish these measurements, test pieces of BaF₂ and ZnSe are placed in a small oven that is located in one leg of the 10.59/5.22 micron two-color interferometer TIP prototype. The oven temperature was varied from room temperature to ∼90 °C, which is slightly higher than the expected window temperatures on ITER. The vibration compensated phase shifts measured for these two materials are dϕᵥ꜀ /dT/L = 0.0121 and −0.275 deg. /°C/cm −1 for BaF₂ and ZnSe, respectively. From these measurements, it was concluded that temperature variations of the environmental/secondary confinement barrier window and primary vacuum window have to be suppressed within approximately 1°C and 2°C, respectively, in order to passively meet the electron density measurement requirements in ITER.

Akiyama, T (ORCID:0000000198469795)↗

Rotation and neoclassical ripple transport in ITER

Neoclassical transport in the presence of non-axisymmetric magnetic fields causes a toroidal torque known as neoclassical toroidal viscosity (NTV). The toroidal symmetry of ITER will be broken by the finite number of toroidal field coils and by test blanket modules (TBMs). The addition of ferritic inserts (FIs) will decrease the magnitude of the toroidal field ripple. 3D magnetic equilibria in the presence of toroidal field ripple and ferromagnetic structures are calculated for an ITER steady-state scenario using the Variational Moments Equilibrium Code (VMEC). Furthermore, neoclassical transport quantities in the presence of these error fields are calculated using the Stellarator Fokker-Planck Iterative Neoclassical Conservative Solver (SFINCS).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Advances in the steady-state hybrid regime in DIII-D – a fully non-inductive, ELM-suppressed scenario for ITER

Here, the hybrid regime with beta, collisionality, safety factor and plasma shape relevant to the ITER steady-state mission has been successfully integrated with ELM suppression by applying an odd parity n=3 resonant magnetic perturbation (RMP). Fully non-inductive hybrids in the DIII-D tokamak with high beta (< β> ≤ 2.8%) and high confinement (H 98y2 ≤ 1.4) in the ITER similar shape have achieved zero surface loop voltage for up to two current relaxation times using efficient central current drive from ECCD and NBCD. The n=3 RMP causes surprisingly little increase in thermal transport during ELM suppression. Poloidal magnetic flux pumping in hybrid plasmas maintains q above 1 without loss of current drive efficiency, except that experiments show that extremely peaked ECCD profiles can create sawteeth. During ECCD, Alfvén eigenmode (AE) activity is replaced by a more benign fishbone-like mode, reducing anomalous beam ion diffusion by a factor of 2. While the electron and ion thermal diffusivities substantially increase with higher ECCD power, the loss of confinement can be offset by the decreased fast ion transport resulting from AE suppression. Extrapolations from DIII-D along a dimensionless parameter scaling path as well as those using self-consistent theory-based modeling show that these ELM-suppressed, fully non-inductive hybrids can achieve the Q fus = 5 ITER steady-state mission.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Use of reconstructed 3D equilibria to determine onset conditions of helical cores in tokamaks for extrapolation to ITER

The threshold for spontaneous symmetry breaking, which results in growth of m/n = 1/1 helical cores, in tokamaks is determined using VMEC. This model is based on a DIII-D hybrid discharge with helical core, and it predicts ITER (15MA scenario) to operate far in the helical core formation regime. Helical cores can only exist in tokamak discharges with monotonic but low, or reversed q-shear and qmin 1 in the core. The helical core is a saturated internal kink mode; its onset threshold is proportional to (dp/d)/B 2 t around q = 1. Below the threshold, applied 3-D fields can drive a helical core to finite size, as in DIIID. Above it, a random 3-D kick causes a bifurcation from axisymmetry and excites a spontaneous helical core. Helical cores occur frequently in C-Mod due to impurities; modeling shows the axisymmetric equilibrium becomes unstable at the end of the current ramp up due to the slow current penetration and a resultant reversed shear q-profile, which lowers the threshold. Lastly, a comparison of the helical core onset threshold for discharges from DIII-D, C-Mod and ITER confirms that while DIII-D is marginally stable, C-Mod and especially ITER are highly susceptible to helical core formation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Assessment of the baseline scenario at q 95 ~ 3 for ITER

The International Tokamak Physics Activity topical group on integrated operational scenarios has compiled a database of stationary H-mode discharges at q 95 ~ 3 from AUG, C-Mod, DIII-D, JET and JT-60U, for both carbon wall and high-Z metal wall experiments with ~3300 entries. The analyses focus on discharges that are stationary for ≥5 thermal energy confinement times to evaluate the baseline scenario proposed for ITER at 15 MA for achieving its goals of Q = 10, fusion power of 500 MW at normalized pressure, β N = 1.8 and normalized confinement as predicted by the standard H-mode scaling, H 98y2 = 1. With the data restricted to stationary H-modes at q 95 ~ 3, the database shows significant variation of thermal energy confinement compared to the standard H-mode scaling (IPB98(y,2)) in dimensionless form. The data show similar scaling with normalized gyro-radius, but more favorable scaling towards lower collision frequency and more favorable scaling with plasma beta. Using all the engineering variables employed in IPB98(y,2), results in an overfit due to correlations among the data. Furthermore, there are significant residual trends in the confinement for plasma current, device size, loss power, and in particular for the plasma density. Significant differences between results obtained for devices with a carbon wall and high-Z metal wall are observed in the data, with data from carbon wall devices providing a larger operating space, encompassing ITER parameters or even exceeding them. H-modes in high-Z metal wall devices have, so-far, not accessed conditions at low collision frequencies, have lower normalized confinement (H 98y2 ~ 0.8–0.9) at low input power or beta, achieving H 98y2 ~ 1.0 only at input powers two times the L- to H-mode transition scaling predictions and at β N ~ 2.0. Hence, only the best H-modes with high-Z metal walls reach ITER baseline performance requirements.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Nonlinear burn control in ITER using adaptive allocation of actuators with uncertain dynamics

Abstract ITER will be the first tokamak to sustain a fusion-producing, or burning, plasma. If the plasma temperature were to inadvertently rise in this burning regime, the positive correlation between temperature and the fusion reaction rate would establish a destabilizing positive feedback loop. Careful regulation of the plasma’s temperature and density, or burn control, is required to prevent these potentially reactor-damaging thermal excursions, neutralize disturbances and improve performance. In this work, a Lyapunov-based burn controller is designed using a full zero-dimensional nonlinear model. An adaptive estimator manages destabilizing uncertainties in the plasma confinement properties and the particle recycling conditions (caused by plasma–wall interactions). The controller regulates the plasma density with requests for deuterium and tritium particle injections. In ITER-like plasmas, the fusion-born alpha particles will primarily heat the plasma electrons, resulting in different electron and ion temperatures in the core. By considering separate response models for the electron and ion energies, the proposed controller can independently regulate the electron and ion temperatures by requesting that different amounts of auxiliary power be delivered to the electrons and ions. These two commands for a specific control effort (electron and ion heating) are sent to an actuator allocation module that optimally maps them to the heating actuators available to ITER: an electron cyclotron heating system (20 MW), an ion cyclotron heating system (20 MW), and two neutral beam injectors (16.5 MW each). Two different actuator allocators are presented in this work. The first actuator allocator finds the optimal mapping by solving a convex quadratic program that includes actuator saturation and rate limits. It is nonadaptive and assumes that the mapping between the commanded control efforts and the allocated actuators (i.e. the effector model) contains no uncertainties. The second actuator allocation module has an adaptive estimator to handle uncertainties in the effector model. This uncertainty includes actuator efficiencies, the fractions of neutral beam heating that are deposited into the plasma electrons and ions, and the tritium concentration of the fueling pellets. Furthermore, the adaptive allocator considers actuator dynamics (actuation lag) that contain uncertainty. This adaptive allocation algorithm is more computationally efficient than the aforementioned nonadaptive allocator because it is computed using dynamic update laws so that finding the solution to a static optimization problem is not required at every time step. A simulation study assesses the performance of the proposed adaptive burn controller augmented with each of the actuator allocation modules.

Physics↗

Toward holistic understanding of the ITER-like resonant magnetic perturbation (RMP) ELM control on KSTAR

Here, KSTAR has clarified a set of unresolved 3D physics issues utilizing the ITER-like in-vessel, three-row, resonant magnetic perturbation (RMP) configurations. Since RMP-driven, edge-localized-modes (ELMs)-crash control elevates the divertor heat flux peak through its impact on edge plasma parameters and transport, a series of intentionally misaligned RMP configurations (IMCs) have been explored to investigate the relationship between RMP ELM control and divertor heat fluxes, while searching for an ideal IMC that could be favorable in both aspects. First of all, the contrasting influence of kink vs anti-kink phasing on the ELM-crash suppression has been articulated, demonstrating the synergistic benefit of 'kink' phasing on ELM-crash-suppression. On the other hand, the three-row IMC in the anti-kink phasing becomes more insensitive to the ELM-crashes at the sub-marginal level of RMP, consistent with theory. Meanwhile, the divertor 'wetted' area of ELM-crash-suppression gets narrower than that of ELM-crash-mitigation, suggesting that ELM-crash-mitigation remains advantageous over ELM-crash-suppression in terms of time-averaged divertor thermal loading. In comparison, based on a set of two-row IMCs, no evidence of divertor heat flux broadening was found during ELM-crash-suppression, supporting a hypothesis that the dispersal of the divertor heat flux in three-row IMCs cannot be driven by helically structured two-row RMPs alone. Among ITER-like three-rows, lower two-row RMPs have been found to be much more effective in suppressing the ELM-crashes than upper two-row RMPs. Although it is quite preliminary, the up/down asymmetric dependence of RMP coupling may be generically attributed to lower-single-null plasmas. Such a holistic understanding of RMP-driven, ELM-crash-control in KSTAR is expected not only to elucidate various subtle points in the vicinity of ELM-crash-suppression, but also to clarify the relevant divertor thermal loading issues for ITER and beyond.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗