Total Plasma Current Research Articles

Control-Oriented Free-Boundary Equilibrium Solver for Tokamaks

A free-boundary equilibrium solver for an axisymmetric tokamak geometry was developed based on the finite difference method and Picard iteration in a rectangular computational area. The solver can run either in forward mode, where external coil currents are prescribed until the converged magnetic flux function ψ(R,Z) map is achieved, or in inverse mode, where the desired plasma boundary, with or without an X-point, is prescribed to determine the required coil currents. The equilibrium solutions are made consistent with prescribed plasma parameters, such as the total plasma current, poloidal beta, or safety factor at a specified flux surface. To verify the mathematical correctness and accuracy of the solver, the solution obtained using this numerical solver was compared with that from an analytic fixed-boundary equilibrium solver based on the EAST geometry. Additionally, the proposed solver was benchmarked against another numerical solver based on the finite-element and Newton-iteration methods in a triangular-based mesh. Finally, the proposed solver was compared with equilibrium reconstruction results in DIII-D experiments.

Formation and termination of runaway beams during vertical displacement events in tokamak disruptions

A simple 0D model which mimics the plasma surrounded by the conducting structures (Kiramov and Breizman 2017 Phys. Plasmas 24 100702) and including self-consistently the vertical plasma motion and the generation of runaway electrons during the disruption is used for an assessment of the effect of vertical displacement events on the runaway current formation and termination. The total plasma current and runaway current at the time the plasma hits the wall is estimated and the effect of injecting impurities into the plasma is evaluated. In the case of ITER, with a highly conducting wall, although the total plasma current when the plasma touches the wall is the same for any number of injected impurities, however the fraction of the plasma current carried by runaway electrons can significantly decrease for large enough amounts of impurities. The plasma velocity is larger and the time when the plasma hits the wall shorter for lower runaway currents, which are obtained when larger amounts of impurities are injected. When the plasma reaches the wall, the scraping-off of the runaway beam occurs and the current is terminated. During this phase, the plasma vertical displacement velocity and electric field can substantially increase leading to the deposition of a noticeable amount of energy on the runaway electrons (∼hundreds of MJ). It is found that an early second impurity injection reduces somewhat the amount of energy deposited by the runaways. Also larger temperatures of the companion plasma during the scraping-off might be efficient in reducing the power fluxes due to the runaways onto the PFCs. The plasma reaches the q a = 2 limit before the runaway electron current is terminated and by that time the amount of energy deposited on the runaway electrons can be substantially lower than that expected until the beam is fully terminated. Negligible additional conversion of magnetic into runaway kinetic energy is predicted during the runaway deconfinement following the large magnetic fluctuations after q a = 2 is crossed for characteristic deconfinement times lower than 0.1 ms which is a characteristic timescale for ideal MHD instabilities to develop.

Final design of the Fiber-Optic Current Sensor bundle in the ITER buildings

The Fiber-Optic Current Sensor (FOCS) will be used in ITER to measure the total plasma current using Faraday effect. Sensing spun fibers will be placed around the Vacuum Vessel (VV) on a poloidal loop in two different VV sectors. To link the loop fibers with the reading instruments, placed several tens of meters away in electrical cabinets, a fiber bundle link is needed. This fiber bundle would start from the tokamak building port-cell, cross several rooms and end in the ITER diagnostics building. The design of such bundle needs to fulfil the unique requirements of ITER buildings, some of them related to safety, such as non-propagation of smoke and fire, as well as maximum allowable leak rate, between two separate rooms. To this regard, the FOCS fiber bundle is classified as a Safety-Relevant (SR) component. To fulfil these requirements, and at the same time achieve a good design using as many Commercial Off The Shelf (COTS) components as possible, the fiber bundle design entailed a complete qualification procedure, aimed at benchmarking the use of the selected components against the required constraints. This paper presents the final design and qualification of the FOCS fiber bundle, from the applicable requirements to the final technical solution, benchmarked through careful qualification tests. Being the first work addressing qualification of safety-relevant fiber bundles in ITER buildings, this work is also considered to be relevant for current and future diagnostics in ITER using similar components.

Calculation of Coil Currents for Equilibria with Reversed Plasma Current Density in J-Text Tokamak

To design the alternating current operation of the J-text tokamak, this paper studies the plasma equilibrium during the plasma toroidal current reversing. First, four alternative current reversal equilibrium configurations with circular cross section when the total toroidal current becomes zero were constructed. Then, the profile of the plasma current was used as the input of a fixed boundary equilibrium solver to calculate the poloidal field coil currents. According to the results, the relationship between the plasma current distribution and currents in ohmic heating coils and vertical field coils is discussed. Finally, the plasma equlibrium of a toroidal current reversal with total plasma current from 40 kA to –40 kA in 40 ms was designed.

Rapid optimization of stationary tokamak plasmas in RAPTOR: demonstration for the ITER hybrid scenario with neural network surrogate transport model QLKNN

This work presents a fast and robust method for optimizing the stationary radial distribution of temperature, density and parallel current density in a tokamak plasma and its application to first-principle-based modeling of the ITER hybrid scenario. A new solver is implemented in the RAPTOR transport code, enabling direct evaluation of the stationary solution to which the radial plasma profiles evolve. Coupled to a neural network emulation of the quasi-linear gyrokinetic QuaLiKiz transport model (QLKNN-hyper-10D), a first-principle-based estimate of the stationary state of the core plasma can be found at unprecedented computational speed (typically a few seconds on standard hardware). The stationary state solver is then embedded in a numerical optimization scheme, allowing the optimization of tokamak plasma scenarios in only a few minutes. The proposed method is applied to investigate the performance of ITER hybrid scenarios at different values of total plasma current, plasma density and pedestal height and for different power contributions in a heating mix consisting of electron cyclotron and neutral beam heating. Optimizing the radial distribution of electron cyclotron current drive (ECCD) deposition, the q profile is tailored to maximize the fusion gain Q, by maximizing the energy confinement predicted through the first-principles-based transport model, while satisfying q > 1, avoiding sawtooth oscillations. It is found that optimal use of ECCD in ITER hybrid scenarios is to deposit power as close to the core as possible, while maintaining sufficient off-axis current drive to keep q above 1. Upper limits for the fusion gain Q are shown to be constrained either by minimum power requirements for the separatrix power flow to maintain H-mode or by minimum current drive requirements for q profile tailoring. Finally, it is shown that the ITER hybrid scenario operating window is significantly extended by an upgrade of the electron cyclotron power to 40 MW.

Resistive versus ideal plasma response to RMP fields in DIII-D: roles of q95 and X-point geometry

Toroidal computation of the plasma response to the n = 2 (n is the toroidal mode number) resonant magnetic perturbation field, based on an H-mode plasma in DIII-D, is carried out for the purpose of investigating the role of the ideal versus resistive plasma response models while scanning the plasma safety factor (q) at the edge. Both response models, implemented in the MARS-F code (Liu et al 2000 Phys. Plasmas 7 3681), show significant amplification of the kink-peeling response in certain q-windows. A longstanding issue addressed in this work is the sensitivity of the q-window versus smoothing of the X-point geometry of the plasma separatrix. For this purpose, scan of the safety factor in 2D space (q95, qa) is carried out, where the q-value at 95% of the equilibrium poloidal flux, q95, is scanned by varying the total plasma current, whilst the edge safety factor qa is varied by gradually smoothing the X-point geometry at fixed total plasma current. Transition to the edge-peeling amplification domain is well described by simple analytic curves relating q95 and qa in the (q95, qa) space. These analytic curves not only help to quantify the sensitivity of the q95-window versus qa variation, but also establish the q95 window in the asymptotic limit of infinite qa (i.e. in the presence of true X-points). The ‘resonant’ q95 values (3.6, 4.05, 4.5) are found to roughly exhibit 1/n periodicity at infinite qa. Both ideal and resistive plasma response models, as well as the computed field at different pick-up locations, yield the same analytic curves describing transition to strong edge-peeling response. Detailed analysis of the computed plasma response and comparison with experiments are also performed.

Induction and stabilization of neoclassical tearing modes on HL-2A tokamak

The induction and stabilization characteristics of neoclassical tearing modes (NTMs) on Hl-2A tokamak are presented. NTMs are induced by sawtooth oscillations of the m/n = 1/1 mode via toroidal coupling and by gas puffing fuelling in the experiments of HL-2A tokamak. The evolutions of NTMs are modeled with a two dimensional non-linear cylindrical tearing mode code and compared with the experimental observations. The stabilization of NTMs with electron cyclotron current drive (ECCD) is numerically studied. It is found that a driven current as low as 0.015 of total plasma current can stabilize or even fully suppress NTMs. On the other hand, a small radial misalignment decreases the stabilizing effect considerably. In addition, a phase misalignment blocks the NTM stabilization and results in island width saturation at a moderate value. The effects of island rotation frequency on NTM stabilization are explored. Furthermore, it is found that the stabilizing effect can be improved efficiently by applying ECCD when the island width is less than a threshold value.

Robust control of the current profile and plasma energy in EAST

Integrated control of the toroidal current density profile, or alternatively the q-profile, and plasma stored energy is essential to achieve advanced plasma scenarios characterized by high plasma confinement, magnetohydrodynamics stability, and noninductively driven plasma current. The q-profile evolution is closely related to the evolution of the poloidal magnetic flux profile, whose dynamics is modeled by a nonlinear partial differential equation (PDE) referred to as the magnetic-flux diffusion equation (MDE). The MDE prediction depends heavily on the chosen models for the electron temperature, plasma resistivity, and non-inductive current drives. To aid control synthesis, control-oriented models for these plasma quantities are necessary to make the problem tractable. However, a relatively large deviation between the predictions by these control-oriented models and experimental data is not uncommon. For this reason, the electron temperature, plasma resistivity, and non-inductive current drives are modeled for control synthesis in this work as the product of an “uncertain” reference profile and a nonlinear function of the different auxiliary heating and current-drive (H&CD) source powers and the total plasma current. The uncertainties are quantified in such a way that the family of models arising from the modeling process is able to capture the q-profile and plasma stored energy dynamics from a typical EAST shot. A control-oriented nonlinear PDE model is developed by combining the MDE with the “uncertain” models for the electron temperature, plasma resistivity, and non-inductive current drives. This model is then rewritten into a control framework to design a controller that is robust against the modeled uncertainties. The resulting controller utilizes EAST's H&CD powers and total plasma current to regulate the q profile and plasma stored energy even when mismatches between modeled and actual dynamics are present. The effectiveness of the controller is demonstrated through nonlinear simulations.

Plasma shape optimization for EAST tokamak using orthogonal method**Project supported by the National Natural Science Foundation of China (Grant No. 51677051) and the Institute of Plasma Physics, Chinese Academy of Sciences.

It is necessary to reduce the currents of poloidal field (PF) coils as small as possible, during the static equilibrium design procedure of Experimental Advanced Superconductive Tokamak (EAST). The quasi-snowflake (QSF) divertor configuration is studied in this paper. Starting from a standard QSF plasma equilibrium, a new QSF equilibrium with 300 kA total plasma current is designed. In order to reduce the currents of PF6 and PF14, the influence of plasma shape on PF coil current distribution is analyzed. A fixed boundary equilibrium solver based on a non-rigid plasma model is used to calculate the flux distribution and PF coil current distribution. Then the plasma shape parameters are studied by the orthogonal method. According to the result, the plasma shape is redefined, and the calculated equilibrium shows that the currents of PF6 and PF14 are reduced by 3.592 kA and 2.773 kA, respectively.

Modelling and experimental validation of RFX-mod Tokamak shaped discharges

A general procedure for computing plasma equilibria through the CREATE-L code has been applied to RFX-mod shaped tokamak discharges in order to produce very accurate models with respect to experimental data. It involves the solution of a constrained non-linear minimization problem to estimate the CREATE-L free parameters by using an iterative scheme trying to minimize the discrepancy between the poloidal magnetic field experimentally measured by pick-up coils and the one simulated by the model. The procedure has been applied considering eleven experimental shots spanning the whole range of poloidal beta achieved in the RFX-mod tokamak. A preliminary sensitivity analysis showed a non-negligible dependence of static equilibria on variations of the total plasma current with respect to the measurement provided by Rogowski coils. Thus, the total plasma current has been set as an additional degree of freedom in the minimization problem assuming values between the Rogowski measurement and the value of the discrete line integral of the poloidal magnetic field measured by the pick-up coils. In all cases under analysis, the iterative procedure showed that the most accurate equilibrium is obtained with a plasma current higher than the Rogowski measurement but lower than the pick-up coils line integral.

A power-balance model for local helicity injection startup in a spherical tokamak

A 0D circuit model for predicting in Local Helicity Injection (LHI) discharges is developed. Analytic formulas for estimating the surface flux of finite- plasmas developed by Hirshman and Neilson (1986 Phys. Fluids 29 790) are modified and expanded to treat highly shaped, ultralow- tokamak geometry using a database of representative equilibria. Model predictions are compared to sample LHI discharges in the Pegasus spherical tokamak, and are found to agree within 15% of experimental . High performance LHI discharges are found to follow the Taylor relaxation current limit for approximately the first half of the current ramp, or 75 kA. The second half of the current ramp follows a limit imposed by power-balance as plasmas expand from high-A to ultralow-A. This shape evolution generates a significant drop in external plasma inductance, effectively using the plasma’s initially high inductance to drive the current ramp and provide >70% of the current drive V-s. Projections using this model indicate the relative influences of higher helicity input rate and injector current on the attainable total plasma current.

Stability analysis of ELMs in long-pulse discharges with ELITE code on EAST tokamak

One challenge in long-pulse and high performance tokamak operation is to control the edge localized modes (ELMs) to reduce the transient heat load on plasma facing components. Minute-scale discharges in H-mode have been achieved repeatedly on Experimental Advanced Superconducting Tokamak (EAST) since the 2016 campaign and understanding the characteristics of the ELMs in these discharges can be helpful for effective ELM control in long-pulse discharges. The kinetic profile diagnostics recently developed on EAST make it possible to perform the pedestal stability analysis quantitatively. Pedestal stability calculation of a typical long-pulse discharge with ELITE code is presented. The ideal linear stability results show that the ELM is dominated by toroidal mode number n around 10–15 and the most unstable mode structure is mainly localized in the steep pressure gradient region, which is consistent with experimental results. Compared with a typical type-I ELM discharge with larger total plasma current (Ip = 600 kA), pedestal in the long-pulse H-mode discharge (Ip = 450 kA) is more stable in peeling-ballooning instability and its critical peak pressure gradient is evaluated to be 65% of the former. Two important features of EAST tokamak in the long-pulse discharge are presented by comparison with other tokamaks, including a wider pedestal correlated with the poloidal pedestal beta and a smaller inverse aspect ratio and their effects on the pedestal stability are discussed. The effects of uncertainties in measurements on the linear stability results are also analyzed, including the edge electron density profile position, the separatrix position and the line-averaged effective ion charge value.

Comparison of runaway electron generation parameters in small, medium-sized and large tokamaks—A survey of experiments in COMPASS, TCV, ASDEX-Upgrade and JET

This paper presents a survey of the experiments on runaway electrons (RE) carried out recently in frames of EUROFusion Consortium in different tokamaks: COMPASS, ASDEX-Upgrade, TCV and JET. Massive gas injection (MGI) has been used in different scenarios for RE generation in small and medium-sized tokamaks to elaborate the most efficient and reliable ones for future RE experiments. New data on RE generated at disruptions in COMPASS and ASDEX-Upgrade was collected and added to the JET database. Different accessible parameters of disruptions, such as current quench rate, conversion rate of plasma current into runaways, etc have been analysed for each tokamak and compared to JET data. It was shown, that tokamaks with larger geometrical sizes provide the wider limits for spatial and temporal variation of plasma parameters during disruptions, thus extending the parameter space for RE generation. The second part of experiments was dedicated to study of RE generation in stationary discharges in COMPASS, TCV and JET. Injection of Ne/Ar have been used to mock-up the JET MGI runaway suppression experiments. Secondary RE avalanching was identified and quantified for the first time in the TCV tokamak in RE generating discharges after massive Ne injection. Simulations of the primary RE generation and secondary avalanching dynamics in stationary discharges has demonstrated that RE current fraction created via avalanching could achieve up to 70–75% of the total plasma current in TCV. Relaxations which are reminiscent the phenomena associated to the kinetic instability driven by RE have been detected in RE discharges in TCV. Macroscopic parameters of RE dominating discharges in TCV before and after onset of the instability fit well to the empirical instability criterion, which was established in the early tokamaks and examined by results of recent numerical simulations.

Influence of helical external driven current on nonlinear resistive tearing mode evolution and saturation in tokamaks

The influences of helical driven currents on nonlinear resistive tearing mode evolution and saturation are studied by using a three-dimensional toroidal resistive magnetohydrodynamic code (CLT). We carried out three types of helical driven currents: stationary, time-dependent amplitude, and thickness. It is found that the helical driven current is much more efficient than the Gaussian driven current used in our previous study [S. Wang et al., Phys. Plasmas 23(5), 052503 (2016)]. The stationary helical driven current cannot persistently control tearing mode instabilities. For the time-dependent helical driven current with fcd=0.01 and δcd<0.04, the island size can be reduced to its saturated level that is about one third of the initial island size. However, if the total driven current increases to about 7% of the total plasma current, tearing mode instabilities will rebound again due to the excitation of the triple tearing mode. For the helical driven current with time dependent strength and thickness, the reduction speed of the radial perturbation component of the magnetic field increases with an increase in the driven current and then saturates at a quite low level. The tearing mode is always controlled even for a large driven current.

Model-based optimal scenario planning in EAST

Ongoing work in the fusion community focuses on developing advanced plasma scenarios characterized by high plasma confinement, magnetohydrodynamic (MHD) stability, and noninductively driven plasma current. The toroidal current density profile, or alternatively the q profile, together with the normalized beta, are often used to characterize these advanced scenarios. The development of these advanced scenarios is experimentally carried out by specifying the devices’ actuator trajectory waveforms, such as the total plasma current, the plasma density, and the auxiliary heating and current-drive (H&CD) sources based on trial-and-error basis. In this work, a model-based numerical optimization approach is followed to complement the experimental effort on actuator trajectory planning in the EAST tokamak. The evolution of the q profile is closely related to the evolution of the poloidal magnetic flux profile, whose dynamics is modeled by a nonlinear partial differential equation (PDE) referred to as the magnetic-flux diffusion equation (MDE). In this work, the MDE is combined with physics-based correlations obtained from EAST experimental data for the plasma density, temperature, resistivity and non-inductive current drives to develop a control-oriented nonlinear PDE model. The optimization objective is to design feedforward trajectories for the plasma current, plasma density, electron cyclotron heating power, neutral beam injection power and lower hybrid current drive power that steer the plasma to desired q profile and βN such that the achieved state is stationary in time. The optimization is subject to the plasma dynamics (described by the physics-based PDE model) and plasma state and actuator constraints, such as the maximum available amount of H&CD power and MHD stability limits. This defines a nonlinear, constrained optimization problem that is solved by employing sequential quadratic programming. The optimized actuator trajectories are assessed in nonlinear transport simulations in preparation for experimental tests in EAST.

Effect of the Shafranov shift and the gradient of β on intrinsic momentum transport in up–down asymmetric tokamaks

Tokamaks with up–down asymmetric poloidal cross-sections spontaneously rotate due to turbulent transport of momentum. In this work, we investigate the effect of the Shafranov shift on this intrinsic rotation, primarily by analyzing tokamaks with tilted elliptical flux surfaces. By expanding the Grad–Shafranov equation in the large aspect ratio limit we calculate the magnitude and direction of the Shafranov shift in tilted elliptical tokamaks. The results show that, while the Shafranov shift becomes up–down asymmetric and depends strongly on the tilt angle of the flux surfaces, it is insensitive to the shape of the current and pressure profiles (when the geometry, total plasma current, and average pressure gradient are kept fixed). Next, local nonlinear gyrokinetic simulations of these MHD equilibria are performed with GS2, which reveal that the Shafranov shift can significantly enhance the momentum transport. However, to be consistent, the effect of (i.e. the radial gradient of β) on the magnetic equilibrium was also included, which was found to significantly reduce momentum transport. Including these two competing effects broadens the rotation profile, but leaves the on-axis value of the rotation roughly unchanged. Consequently, the shape of the β profile has a significant effect on the rotation profile of an up–down asymmetric tokamak.

Neutron emission in neutral beam heated KSTAR plasmas and its application to neutron radiography

The main mission of Korea Superconducting Tokamak Advanced Research (KSTAR) program is exploring the physics and technologies of high performance steady state Tokamak operation that are essential for ITER and fusion reactor. Since the successful first operation in 2008, the plasma performance is enhanced and duration of H-mode is extended to around 50s which corresponds to a few times of current diffusion time and surpassing the current conventional Tokamak operation. In addition to long-pulse operation, the operational boundary of the H-mode discharge is further extended over MHD no-wall limit(βN∼4) transiently and higher stored energy region is obtained by increased total heating power (∼6MW) and plasma current (Ip up to 1MA for ∼10s). Heating system consists of various mixtures (NB, ECH, LHCD, ICRF) but the major horse heating resource is the neutral beam(NB) of 100keV with 4.5MW and most of experiments are conducted with NB. So there is a lot of production of fast neutrons coming from via D(d,n)3He reaction and it is found that most of neutrons are coming from deuterium beam plasma interaction. Nominal neutron yield and the area of beam port is about 1013–1014/s and 1m2 at the closest access position of the sample respectively and neutron emission could be modulated for application to the neutron radiography by varying NB power. This work reports on the results of quantitative analysis of neutron emission measurements and results are discussed in terms of beam-plasma interaction and plasma confinement. It also includes the feasibility study of neutron radiography using KSTAR.

Backstepping Control of the Toroidal Plasma Current Profile in the DIII-D Tokamak

One of the most promising devices for realizing power production through nuclear fusion is the tokamak. To maximize performance, it is preferable that tokamak reactors achieve advanced operating scenarios characterized by good plasma confinement, improved magnetohydrodynamic stability, and a largely noninductively driven plasma current. Such scenarios could enable steady-state reactor operation with high fusion gain, the ratio of produced fusion power to the external power provided through the plasma boundary. For certain advanced scenarios, control of the spatial profile of the plasma current will be essential. The complexity of the current profile dynamics, arising due to nonlinearities and couplings with many other plasma parameters, motivates the use of model-based control algorithms that can account for the system dynamics. A first-principles-driven, control-oriented model of the current profile evolution in low-confinement mode (L-mode) discharges in the DIII-D tokamak is employed to address the problem of regulating the current profile evolution around desired trajectories. In the primarily inductive L-mode discharges considered in this paper, the boundary condition, which is dependent on the total plasma current, has the largest influence on the current profile dynamics, motivating the design of a boundary feedback control law to improve the system performance. The backstepping control design technique provides a systematic method to obtain a boundary feedback law through the transformation of a spatially discretized version of the original system into an asymptotically stable target system with desirable properties. Through a nonlinear transformation of the available physical actuators, the resulting control scheme produces references for the total plasma current, total power, and line averaged density, which are tracked by existing dedicated control loops. Adaptiveness is added to the control scheme to improve upon the backstepping controller's disturbance rejection and tracking capability. Prior to experimental testing, a Simserver simulation was carried out to study the controller's performance and ensure proper implementation in the DIII-D Plasma Control System. An experimental test was performed on DIII-D to test the ability of the controller to reject input disturbances and perturbations in initial conditions and to demonstrate the feasibility of the proposed control approach.

Fibrillation atriale

Active closed-loop control of the plasma safety factor profile (q-profile) and internal energy dynamics in nuclear fusion tokamak devices has the potential to significantly impact the success of the ITER project. These plasma properties are related to both the stability and performance of a given plasma operating scenario. In this work, we develop integrated feedback control algorithms to control the q-profile and internal energy dynamics in DIII-D advanced tokamak (high performance) scenarios. The feedback controllers are synthesized by embedding a nonlinear, physics-based, control-oriented partial differential equation model of the plasma dynamics into the control design and to be robust to uncertainties in the plasma electron density, electron temperature, and plasma resistivity profiles. The auxiliary heating and current-drive system and the total plasma current are the actuators utilized by the feedback controllers to control the plasma dynamics. Finally, the feedback controllers are tested both through simulations based on the physics-based model and experimentally in the DIII-D tokamak.

First-Principles-Driven Model-Based Control of the Poloidal Magnetic Flux Profile at the DIII-D Tokamak

Abstract Efficient, high-gain operation of a tokamak device requires the achievement of certain radial shapes for the toroidal current profile. The evolution in time of the toroidal current profile in tokamaks is related to the evolution of the poloidal magnetic flux profile. A model-based control approach for the regulation of the poloidal magnetic flux profile at the DIII-D tokamak is proposed in this work. The model describing the poloidal flux evolution is based on a control-oriented formulation of the magnetic diffusion equation. Auxiliary heating and current drive (H&CD) systems including electron cyclotron (EC) and neutral beam injection (NBI) along with the total plasma current are used as actuators to manipulate the profile shape. Optimal state feedback control with integral action is used to design a controller to regulate the profile around a target while rejecting disturbances. Combining the profile controller with control of the plasma stored energy is found to improve tracking performance. Simulations and experimental results are presented to demonstrate the controller's effectiveness.