Neutral Beam Injection Power Research Articles

Integrated current profile, normalized beta and NTM control in DIII-D

There is an increasing need for integrating individual plasma-control algorithms with the ultimate goal of simultaneously regulating more than one plasma property. Some of these integrated-control solutions should have the capability of arbitrating the authority of the individual plasma-control algorithms over the available actuators within the tokamak. Such decision-making process must run in real time since its outcome depends on the plasma state. Therefore, control architectures including supervisory and/or exception-handling algorithms will play an essential role in future fusion reactors like ITER. However, most plasma-control experiments in present devices have focused so far on demonstrating control solutions for isolated objectives. In this work, initial experimental results are reported for simultaneous current-profile control, normalized-beta control, and Neoclassical Tearing Mode (NTM) suppression in DIII-D. Neutral beam injection (NBI), electron-cyclotron (EC) heating & current drive (H&CD), and plasma current modulation are the actuation methods. The NBI power and plasma current are always modulated by the Profile Control category within the DIII-D Plasma Control System (PCS) in order to control both the current profile and the normalized beta. EC H&CD is utilized by either the Profile Control or the Gyrotron categories within the DIII-D PCS as dictated by the Off-Normal and Fault Response (ONFR) system, which monitors the occurrence of an NTM and regulates the authority over the gyrotrons. The total EC power and poloidal mirror angles are the gyrotron-related actuation variables. When no NTM suppression is required, the gyrotrons are used by the Profile Control category, but when NTM suppression is required, the ONFR transfers the authority over the gyrotrons to the NTM stabilization algorithm located in the Gyrotron category. Initial experimental results show that simultaneous control of different aspects of the plasma dynamics may improve the overall control and plasma performances. Also, the potential of the ONFR system to successfully integrate competing control algorithms is demonstrated.

TRANSP-based optimization towards tokamak scenario development

An optimization approach that incorporates the predictive transport code TRANSP is proposed for tokamak scenario development. Optimization methods are often employed to develop open-loop control strategies to aid access to high performance tokamak scenarios. In general, the optimization approaches use control-oriented models, i.e. models that are reduced in complexity and prediction accuracy as compared to physics-oriented transport codes such as TRANSP. In the presented approach, an optimization procedure using the TRANSP code to simulate the tokamak plasma is considered for improved predictive capabilities. As a test case, the neutral beam injection (NBI) power is optimized to develop a control strategy that maximizes the noninductive current fraction during the ramp-up phase for NSTX-U. Simulation studies towards the achievement of noninductive ramp-up in NSTX-U have already been carried out with the TRANSP code. The optimization-based approach proposed in this work is used to maximize the noninductive current fraction during ramp-up in NSTX-U, demonstrating that the scenario development task can be automated. An additional test case considers optimization of the current ramp rate in DIII-D for obtaining a stationary plasma characterized by a flat loop voltage profile in the flattop phase.

Experimental characterization of Alfvén modes in a field-reversed configuration plasma

High power neutral beam injection into the C-2U advanced field-reversed configuration (FRC) stimulates at least three distinct energetic particle modes, none of which have been linked to a performance limitation. Here we present an experimental characterization of one mode in particular, a high frequency mode () observed in the decaying phase of the FRC. The mode is reminiscent of the compressional Alfvén eigenmodes observed in spherical tokamaks: its frequency scales with the edge Alfvén velocity and it co-propagates with injected beam ions perpendicular to the magnetic field. There are multiple pieces of evidence that the mode resides outside of the FRC and therefore likely has little to no effect on the confinement properties of the FRC itself.

W transport and accumulation control in the termination phase of JET H-mode discharges and implications for ITER

Tokamak operation with W PFCs is associated with specific challenges for impurity control, which may be particularly demanding in the transition from stationary H-mode to L-mode. To address W control issues in this phase, dedicated experiments have been performed at JET including the variation of the decrease of the power and current, gas fuelling and central ion cyclotron heating (ICRH), and applying active ELM control by vertical kicks. The experimental results obtained demonstrate the key role of maintaining ELM control to control the W concentration in the exit phase of H-modes with slow (ITER-like) ramp-down of the neutral beam injection power in JET.For these experiments, integrated fully predictive core+edge+SOL transport modelling studies applying discrete models for the description of transients such as sawteeth and ELMs have been performed for the first time with the JINTRAC suite of codes for the entire transition from stationary H-mode until the time when the plasma would return to L-mode focusing on the W transport behaviour. Simulations have shown that the existing models can appropriately reproduce the plasma profile evolution in the core, edge and SOL as well as W accumulation trends in the termination phase of JET H-mode discharges as function of the applied ICRH and ELM control schemes, substantiating the ambivalent effect of ELMs on W sputtering on one side and on edge transport affecting core W accumulation on the other side. The sensitivity with respect to NB particle and momentum sources has also been analysed and their impact on neoclassical W transport has been found to be crucial to reproduce the observed W accumulation characteristics in JET discharges.In this paper the results of the JET experiments, the comparison with JINTRAC modelling and the adequacy of the models to reproduce the experimental results are described and conclusions are drawn regarding the applicability of these models for the extrapolation of the applied W accumulation control techniques to ITER.

Validation of the kinetic-turbulent-neoclassical theory for edge intrinsic rotation in DIII-D

In a recent kinetic model of edge main-ion (deuterium) toroidal velocity, intrinsic rotation results from neoclassical orbits in an inhomogeneous turbulent field [T. Stoltzfus-Dueck, Phys. Rev. Lett. 108, 065002 (2012)]. This model predicts a value for the toroidal velocity that is co-current for a typical inboard X-point plasma at the core-edge boundary (ρ ∼ 0.9). Using this model, the velocity prediction is tested on the DIII-D tokamak for a database of L-mode and H-mode plasmas with nominally low neutral beam torque, including both signs of plasma current. Values for the flux-surface-averaged main-ion rotation velocity in the database are obtained from the impurity carbon rotation by analytically calculating the main-ion—impurity neoclassical offset. The deuterium rotation obtained in this manner has been validated by direct main-ion measurements for a limited number of cases. Key theoretical parameters of ion temperature and turbulent scale length are varied across a wide range in an experimental database of discharges. Using a characteristic electron temperature scale length as a proxy for a turbulent scale length, the predicted main-ion rotation velocity has a general agreement with the experimental measurements for neutral beam injection (NBI) powers in the range PNBI < 4 MW. At higher NBI power, the experimental rotation is observed to saturate and even degrade compared to theory. TRANSP-NUBEAM simulations performed for the database show that for discharges with nominally balanced—but high powered—NBI, the net injected torque through the edge can exceed 1 Nm in the counter-current direction. The theory model has been extended to compute the rotation degradation from this counter-current NBI torque by solving a reduced momentum evolution equation for the edge and found the revised velocity prediction to be in agreement with experiment. Using the theory modeled—and now tested—velocity to predict the bulk plasma rotation opens up a path to more confidently projecting the confinement and stability in ITER.

Turbulent transport stabilization by ICRH minority fast ions in low rotating JET ILW L-mode plasmas

The first experimental demonstration that fast ion induced stabilization of thermal turbulent transport takes place also at low values of plasma toroidal rotation has been obtained in JET ILW (ITER-like wall) L-mode plasmas with high (3He)-D ICRH (ion cyclotron resonance heating) power. A reduction of the gyro-Bohm normalized ion heat flux and higher values of the normalized ion temperature gradient have been observed at high ICRH power and low NBI (neutral beam injection) power and plasma rotation. Gyrokinetic simulations indicate that ITG (ion temperature gradient) turbulence stabilization induced by the presence of high-energetic 3He ions is the key mechanism in order to explain the experimental observations. Two main mechanisms have been identified to be responsible for the turbulence stabilization: a linear electrostatic wave-fast particle resonance mechanism and a nonlinear electromagnetic mechanism. The dependence of the stabilization on the 3He distribution function has also been studied.

Kinetic equilibrium reconstruction for the NBI- and ICRH-heated H-mode plasma on EAST tokamak

The equilibrium reconstruction is important to study the tokamak plasma physical processes. To analyze the contribution of fast ions to the equilibrium, the kinetic equilibria at two time-slices in a typical H-mode discharge with different auxiliary heatings are reconstructed by using magnetic diagnostics, kinetic diagnostics and TRANSP code. It is found that the fast-ion pressure might be up to one-third of the plasma pressure and the contribution is mainly in the core plasma due to the neutral beam injection power is primarily deposited in the core region. The fast-ion current contributes mainly in the core region while contributes little to the pedestal current. A steep pressure gradient in the pedestal is observed which gives rise to a strong edge current. It is proved that the fast ion effects cannot be ignored and should be considered in the future study of EAST.

High fusion performance at high Ti/Te in JET-ILW baseline plasmas with high NBI heating power and low gas puffing

This paper presents the transport analysis of high density baseline discharges in the 2016 experimental campaign of the Joint European Torus with the ITER-Like Wall (JET-ILW), where a significant increase in the deuterium–deuterium (D–D) fusion neutron rate (~2.8 × 1016 s−1) was achieved with stable high neutral beam injection (NBI) powers of up to 28 MW and low gas puffing. Increase in Ti exceeding Te were produced for the first time in baseline discharges despite the high electron density; this enabled a significant increase in the thermal fusion reaction rate. As a result, the new achieved record in fusion performance was much higher than the previous record in the same heating power baseline discharges, where Ti = Te. In addition to the decreases in collisionality and the increases in ion heating fraction in the discharges with high NBI power, Ti > Te can also be attributed to positive feedback between the high Ti/Te ratio and stabilisation of the turbulent heat flux resulting from the ion temperature gradient driven mode. The high Ti/Te ratio was correlated with high rotation frequency. Among the discharges with identical beam heating power, higher rotation frequencies were observed when particle fuelling was provided by low gas puffing and pellet injection. This reveals that particle fuelling played a key role for achieving high Ti/Te, and the improved fusion performance.

A DT fusion reactor design in field-reversed configuration using normal conductive coils

Assuming continued stability, favorable energy confinement time scaling, and an effective current drive and maintenance methods, a feasible DT fusion reactor design is proposed for field-reversed configuration (FRC) which uses normal conductive copper magnetic field coils at room temperature. The reactor has 3 GW fusion power, 1.5 MW m−2 neutron wall loading at the first wall, and thermal loading less than 1 MW m−2 at diverter plates. Plasma has an almost straight cylindrical shape of 40 m in length and 8.6 m in diameter. FRC can obtain very high beta (over 0.7 in average) and the magnetic field strength of the reactor will be 1.215 T, which can be produced by normal conductive coils having 70 m in length, 17.6 m in diameter, and 1.5 m in thickness with 0.6 effective conductive area ratio. Its Ohmic power loss is ~74 MW, which is less than 10% of the expected electric power output. A scenario to reach ignition from the initial formation is considered. At first, two FRCs are formed at the both ends of the reactor by fast theta-pinch with a negative bias magnetic field 6 m in length and 0.5 m in diameter. The FRCs are accelerated up to 250 km s−1 by the gradient of magnetic field strength towards the center of the burning region, collide with each other, and form a single large FRC. Their kinetic energy is converted to thermal energy, and the merged FRC is 10 m in length and 1.8 m in diameter. This FRC plasma is brought to ignition by intensive neutral beam injection (NBI) heating and particle supply. Given 200 s heating duration, the maximum NBI power is ~250 MW before alpha particle heating becomes significant. After ignition, NBI heating is not required, but there is a possibility that some part of equilibrium current must be supplied by the NBI in the MeV region.

Helium neutral beam injection into ASDEX Upgrade

ASDEX Upgrade's (AUG) neutral beam injection (NBI) system is primarily designed for deuterium injection and delivers 20MW heating power from two injectors with four beams each at 60 and 93keV, respectively. As opposed to the cryo pumps of the JET NBI, the Ti getter pumps of the AUG NBI with a pumping speed of 3×106ℓ/s for hydrogen do not pump helium at all, leaving only the conventional pumping system with <6×103ℓ/s for He. This imposes constraints on the possible operation in helium.In order to prepare for AUG He plasma campaigns, serious trials to operate the AUG NBI with He began in 2014. It was found that despite the lack of high speed pumping up to two beams per injector could be operated simultaneously at reduced feed gas flow without particular restrictions on the beam-on time. For injector 1 the power per beam was limited to 550kW at 40keV by the required filament current in its arc sources, while for injector 2 the limitation came from the bending magnet's power supply that restricted the beam energy to 68keV and the NBI power to 730kW per beam. Thus the maximum available NBI heating power in He amounts to 2.6MW for 10s, the maximum discharge duration of AUG. Helium neutral beam injection into plasma was first tried out in 2014 for two discharges. In 2015 a dedicated He campaign used He NBI in a total of 44 discharges. As He is almost not pumped in the injectors the neutral gas flow into the torus is comparable with the total He gas puff.

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.

Strong suppression of impurity accumulation in steady-state hydrogen discharges with high power NBI heating on LHD

Strong suppression of impurity accumulation is observed in long pulse hydrogen discharges with high power NBI (neutral beam injection) heating (Pnbi > 10 MW) on the large helical device (LHD), even in the impurity accumulation window where the intrinsic impurities such as Fe and C are always accumulated into the plasma core. Density scan experiments in these discharges demonstrate to vanish the window and a new operational regime without impurity accumulation is found in steady state hydrogen discharges. Impurity pinch decreases with increasing ion temperature gradient and carbon Mach number. The peaking of the measured carbon profiles shows strong anti-correlations with the Mach number and its radial gradient. An external torque has a big impact on impurity transport and strong co-current rotation leads to an extremely hollow carbon profile, so-called ‘impurity hole’ observed in high ion temperature modes. Impurity pinch in the plasmas with net zero torque input (balanced NBI injection) is also strongly reduced by increasing ion temperature gradient, which can drive turbulent modes. The combination effect of turbulence and toroidal rotation plays an important role in the impurity transport.

The latest progress of the 1st NBI beamline on HL–2M Tokamak

HL–2M tokamak is under construction in Southwestern Institute of Physics in China, then leading fusion plasma experiment with advanced tokamak configuration will be performed with up to 31MW auxiliary heating power, including 15MW NBI power provided by three neutral beam injection (NBI) systems. The 1st NBI beamline for HL–2M tokamak, employing 4 arc-driven ion sources, which can extract 14.4MW ion beam power, is being developed. The vacuum chamber and main water-cooled components (i.e. ion dumps, calorimeter, deflection magnet, etc.) have been manufactured and assembled. An ion source test bed had been built and testing experiments were carried out and, after testing and optimization, the configuration of ion source was optimized and the experimental maximum extraction parameters for H+ reached 75kV×35A.

Upgrades and application of FIT3D NBI–plasma interaction code in view of LHD deuterium campaigns

This work presents an upgrade of the FIT3D neutral beam–plasma interaction code, part of TASK3D, a transport suite of codes, and its application to LHD experiments in the framework of the preparation for the first deuterium experiments in the LHD. The neutral beam injector (NBI) system will be upgraded to D injection, and efforts have been recently made to extend LHD modelling capabilities to D operations. The implemented upgrades for FIT3D to enable D NBI modelling in D plasmas are presented, with a discussion and benchmark of the models used. In particular, the beam ionization module has been modified and a routine for neutron production estimation has been implemented. The upgraded code is then used to evaluate the NBI power deposition in experiments with different plasma compositions. In the recent LHD campaign, in fact, He experiments have been run to help the prediction of main effects which may be relevant in future LHD D plasmas. Identical H/He experiments showed similar electron density and temperature profiles, while a higher ion temperature with an He majority has been observed. From first applications of the upgraded FIT3D code it turns out that, although more NB power appears to be coupled with the He plasma, the NBI power deposition is unaffected, suggesting that heat deposition does not play a key role in the increased ion temperature with He plasma.

Absolute calibration of neutron detectors on the C-2U advanced beam-driven FRC.

In the C-2U fusion energy experiment, high power neutral beam injection creates a large fast ion population that sustains a field-reversed configuration (FRC) plasma. The diagnosis of the fast ion pressure in these high-performance plasmas is therefore critical, and the measurement of the flux of neutrons from the deuterium-deuterium (D-D) fusion reaction is well suited to the task. Here we describe the absolute, in situ calibration of scintillation neutron detectors via two independent methods: firing deuterium beams into a high density gas target and calibration with a 2 × 107 n/s AmBe source. The practical issues of each method are discussed and the resulting calibration factors are shown to be in good agreement. Finally, the calibration factor is applied to C-2U experimental data where the measured neutron rate is found to exceed the classical expectation.

Modeling and Analysis of Inverter-Type High Voltage Power Supply for NBI Accelerator Grid

The acceleration grid power supply (AGPS) is an important part of the neutral beam injector (NBI) system. With the increase in NBI power, the requirement for output voltage of the AGPS is becoming higher and higher, even reaching to 1000 kV. As the inverter-type high voltage power supply uses inverters located at the lower voltage side to regulate and cut off the output voltage, it is very suitable for high-voltage applications. The proposed scheme for the AGPS of the International Thermonuclear Experimental Reactor (ITER) NBI system comprises five stages, where each stage has an inverter-type power supply with an output voltage of 200 kV. This paper investigates the inverter-type power supply based on neutral point clamped three-phase three-level (TPTL) inverter and three-phase step-up transformer, which is equivalent to an insulated TPTL dc–dc converter connected with a dual dc link. In order to design a preferable controller for the inverter-type power supply, the topology and principles of the TPTL dc–dc converter with duty cycle modulation have been analyzed. It indicates that only when the converter operates at a large duty cycle, the output voltage ripple can be limited to ±5%. Hence, the small signal model of the converter at a duty cycle larger than 2/3 is established. Following the mathematical model, a single-loop PID controller has been designed and a 400 V/6 A prototype has been built. Simulation and experimental results verify the correctness of the theoretical analysis. The system achieves a lower output voltage ripple and fast dynamic response.

Initial operation of the first neutral beamline on EAST

The first neutral beam injector (NBI) system of the Experimental Advanced Superconducting Tokamak (EAST) was installed in the spring of 2014, and the NBI initial experiments were carried out during the EAST operation campaign with its partial design capability. The maximum designed deuterium NB power and energy are 4MW and 80keV, respectively; an extracted beam full energy is mostly below 60keV under the current EAST experimental conditions. The target plasma temperature is obviously increased with the NBI. L–H mode transitions are conducted by only NBI auxiliary heating. The pedestal formation in temperature profiles is observed through Thomson Scattering (TS) diagnostics. The NBI power can be modulated with different frequency and duty ratio has been also well checked.

Determination of tungsten and molybdenum concentrations from an x-ray range spectrum in JET with the ITER-like wall configuration

The and 3p–4d inner shell excitation lines in addition to 2p–3s lines have been identified from the spectrum taken by an upgraded high-resolution x-ray spectrometer. It is found from analysis of the absolute intensities of the and lines that W and Mo concentrations are in the range of and , respectively, with a ratio of ∼5% in JET with the ITER-like wall configuration for ELMy H-mode plasmas with a plasma current of 2.0–2.5 MA, a toroidal magnetic field of 2.7 T and a neutral beam injection power of 14–18 MW. For the purpose of checking self-consistency, it is confirmed that the W concentration determined from the line is in agreement with that from the line within 20% and that the plasma effective charge determined from the continuum of the first order reflection spectrum is also in agreement with that from the second order within 50%. Further, the determined plasma effective charge is in agreement with that determined from a visible spectroscopy, confirming that the sensitivity of the x-ray spectrometer is valid and that the W and the Mo concentrations are also likely to be valid.

Empirical Scaling Laws of Neutral Beam Injection Power in HL-2A Tokamak

We present an experimental method to obtain neutral beam injection (NBI) power scaling laws with operating parameters of the NBI system on HL-2A, including the beam divergence angle, the beam power transmission efficiency, the neutralization efficiency and so on. With the empirical scaling laws, the estimating power can be obtained in every shot of experiment on time, therefore the important parameters such as the energy confinement time can be obtained precisely. The simulation results by the tokamak simulation code (TSC) show that the evolution of the plasma parameters is in good agreement with the experimental results by using the NBI power from the empirical scaling law.

Combined magnetic and kinetic control of advanced tokamak steady state scenarios based on semi-empirical modelling

This paper shows that semi-empirical data-driven models based on a two-time-scale approximation for the magnetic and kinetic control of advanced tokamak (AT) scenarios can be advantageously identified from simulated rather than real data, and used for control design. The method is applied to the combined control of the safety factor profile, q(x), and normalized pressure parameter, βN, using DIII-D parameters and actuators (on-axis co-current neutral beam injection (NBI) power, off-axis co-current NBI power, electron cyclotron current drive power, and ohmic coil). The approximate plasma response model was identified from simulated open-loop data obtained using a rapidly converging plasma transport code, METIS, which includes an MHD equilibrium and current diffusion solver, and combines plasma transport nonlinearity with 0D scaling laws and 1.5D ordinary differential equations. The paper discusses the results of closed-loop METIS simulations, using the near-optimal ARTAEMIS control algorithm (Moreau D et al 2013 Nucl. Fusion 53 063020) for steady state AT operation. With feedforward plus feedback control, the steady state target q-profile and βN are satisfactorily tracked with a time scale of about 10 s, despite large disturbances applied to the feedforward powers and plasma parameters. The robustness of the control algorithm with respect to disturbances of the H&CD actuators and of plasma parameters such as the H-factor, plasma density and effective charge, is also shown.