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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

The Spectrum of Ion Cyclotron Emission from Neutral Beam Injection Heated Plasma on the TUMAN-3M Tokamak

Radiation at the ion cyclotron resonance frequencies in both the ohmic mode and the injection heating mode—ion cyclotron emission (ICE)—is observed on the TUMAN-3M tokamak (minor radius a = 0.25 m, major radius R{sub 0} = 0.5 m, toroidal field B{sub T} = 1 T, plasma current I{sub p} ≤ 180 kA, and average plasma electron density n{sub e} ≤ 5 × 10{sup 19} m{sup –3}). Usually, the ICE is associated with the presence of high-energy ions in the plasma. Such ions may be products of a thermonuclear reaction or arise as a result of additional plasma heating, e.g., by means of electromagnetic waves or by injection of beams of high-energy atoms (neutral beam injection, NBI). The properties of the ICE in the TUMAN-3M tokamak in the NBI-heating mode and their possible connection with the features of fast ion trajectories and their energy spectrum are considered.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Simulation of neutron emission in neutral beam injection heated plasmas with the real-time code RABBIT

In plasmas heated with deuterium beams a deficit of the expected fusion neutron rate is an indicator of the deterioration of the fast-ion confinement, caused, for instance, by magnetohydrodynamic instabilities. The capability of predicting this deficit during the discharge relies on the availability of real-time estimates of the neutron rate from NBI codes which must be fast and accurate at the same time. Therefore, the recently developed real-time RABBIT code for NBI simulations has been extended to output the distribution function and calculate the neutron emission. After the description of this newly installed diagnostics in RABBIT, benchmarks with NUBEAM, a massively used and validated Monte Carlo NBI solver, are discussed on ASDEX-Upgrade and JET cases. A first application for control-room intershot analysis on DIII-D is presented, and the results are compared on a large database with a slower NUBEAM analysis. Further application possibilities, e.g. for real-time control of Alfvén eigenmodes, are outlined.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

On the role of ion temperature gradient turbulence in driving ion thermal transport in neutral beam injection-heated L-mode plasmas in a superconducting tokamak

In this paper, we report a detailed experimental study of the role of ion-temperature-gradient driven turbulence in driving ion thermal transport in neutral beam injection (NBI)-heated L-mode plasmas in Experimental Advanced Superconducting Tokamak (EAST) (Wan et al 2000 Nucl. Fusion 40 1057) coupled with linear and nonlinear gyrokinetic simulations. Significant ion-scale turbulence k ⟂ < cm −1 (k ⟂ < 1.5, where k ⟂ where is the perpendicular wavenumber and ρ s is the ion gyroradius calculated using local electron temperature T e ), measured with a microwave reflectometer, is observed in the plasma core at for r/a ≈ 0.265 and 0.5 (where r is half the diameter of the closed flux surface at a given radial position, and a is half the diameter of the last closed flux surface). Local linear stability analysis with experimental equilibrium quantities at these two radial locations using the GS2 gyrokinetic code shows that the most unstable ion-scale micro-instability is the ion temperature gradient (ITG) mode. Since the computed maximum ion-scale linear growth rates are larger than the local Waltz-Miller E × B shearing rate (Waltz and Miller 1999 Phys. Plasmas 6 4265), the E × B shear is unable to suppress ion-scale turbulence, consistent with the experimental observation of ion-scale turbulence. Experimental ion and electron thermal transports, calculated with power balance analysis, are both anomalous with ion thermal transport being the dominant channel. Nonlinear gyrokinetic simulations using the electrostatic global particle-in-cell gyrokinetic tokamak simulation (GTS) code (Weixing Wang et al 2010 Phys. Plasmas 17 072511) show good quantitative agreement between predicted ion thermal transport and that from the power balance analysis. These results demonstrate for the first time the important role of ITG turbulence in driving thermal transport in NBI-heated L-mode plasmas in EAST, showing the effectiveness of GTS code in simulating electrostatic turbulence in EAST and the need of suppressing ITG turbulence to improve energy confinement in EAST.

Sun, P. J. [Chinese Academy of Sciences (CAS), Hef↗
DOE OSTI2021

Isotope effects on transport in LHD

Abstract Isotope effects are one of the most important issues for predicting future reactor operations. Large helical device (LHD) is the presently working largest stellarator/helical device using super conducting helical coils. In LHD, deuterium experiments started in 2017. Extensive studies regarding isotope effects on transport have been carried out. In this paper, the results of isotope effect studies in LHD are reported. The systematic studies were performed adjusting operational parameters and nondimensional parameters. In L mode like normal confinement plasma, where internal and edge transport barriers are not formed, the scaling of global energy confinement time ( τ E ) with operational parameters shows positive mass dependence ( M 0.27 ; where M is effective ion mass) in electron cyclotron heating plasma and no mass dependence ( M 0.0 ) in neutral beam injection heating plasma. The non-negative ion mass dependence is anti-gyro-Bohm scaling. The role of the turbulence in isotope effects was also found by turbulence measurements and gyrokinetic simulation. Better accessibility to electron and ion internal transport barrier (ITB) plasma is found in deuterium (D) plasma than in hydrogen (H). Gyro kinetic non-linear simulation shows reduced ion heat flux due to the larger generation of zonal flow in deuterium plasma. Peaked carbon density profile plays a prominent role in reducing ion energy transport in ITB plasma. This is evident only in plasma with deuterium ions. New findings on the mixing and non-mixing states of D and H particle transports are reported. In the mixing state, ion particle diffusivities are higher than electron particle diffusivities and D and H ion density profiles are almost identical. In the non-mixing state, ion particle diffusivity is much lower than electron diffusivity. Deuterium and hydrogen ion profiles are clearly different. Different turbulence structures were found in the mixing and non-mixing states suggesting different turbulence modes play a role.

Physics↗
DOE OSTI2015

Tungsten impurity transport experiments in Alcator C-Mod to address high priority research and development for ITER

Experiments in Alcator C-Mod tokamak plasmas in the Enhanced D-alpha H-mode regime with ITER-like mid-radius plasma density peaking and Ion Cyclotron Resonant heating, in which tungsten is introduced by the laser blow-off technique, have shown that accumulation of tungsten in the central region of the plasma does not take place in these conditions. The measurements obtained are consistent with anomalous transport dominating tungsten transport except in the central region of the plasma where tungsten transport is neoclassical, as previously observed in other devices with dominant neutral beam injection heating, such as JET and ASDEX Upgrade. In contrast to such results, yet, the measured scale lengths for plasma temperature and density in the central region of these Alcator C-Mod plasmas, with density profiles relatively flat in the core region due to the lack of core fuelling, are favourable to prevent inter and intra sawtooth tungsten accumulation in this region under dominance of neoclassical transport. Simulations of ITER H-mode plasmas, including both anomalous (modelled by the Gyro-Landau-Fluid code GLF23) and neoclassical transport for main ions and tungsten and with density profiles of similar peaking to those obtained in Alcator C-Mod show that accumulation of tungsten in the central plasma region is also unlikely to occur in stationary ITER H-mode plasmas due to the low fuelling source by the neutral beam injection (injection energy ~ 1 MeV), which is in consistent agreement with findings in the Alcator C-Mod experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Estimation of 14 MeV neutron rate from triton burn-up in future W7-X deuterium plasma campaigns

Fast ion confinement is of major importance for the ignition of a burning fusion plasma. In future deuterium plasma campaigns of the Wendelstein 7-X stellarator, W7-X, the amount of triton burn-up is one possible measure for fast ion confinement. A well-established technique to observe triton burn-up is the 14 MeV neutron rate. In this paper, it is estimated whether an existing scintillating fibre neutron detector is also suited to measure triton burn-up in W7-X with sufficient accuracy. In this work, an estimation is presented, which can be applied to any tokamak or stellarator design and is one-dimensional in the minor radius. The inputs are profiles of density, temperature, and differential volume element as well as the triton slowing-down time. The estimation calculates the thermal deuteron fusion rate and the associated deuteron-triton fusion rate; thus, the triton burn-up generated 14 MeV neutron rate. It neither takes triton diffusion nor explicit losses into account. This thermally generated fusion rate is compared to the neutral beam injection heating induced beam-plasma fusion rate.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Self-induced steady-state magnetic field in the negative ion sources with localized rf power deposition

The study is in the scope of a recent activity on modeling of SPIDER (Source for Production of Ions of Deuterium Extracted from RF plasma) which is under development regarding the neutral beam injection heating system of ITER. The regime of non-ambipolarity in the source, established before, is completed here by introducing in the model the steady state magnetic field, self-induced in the discharge due to the dc current flowing in it. Strong changes in the discharge structure are reported.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Invited Article: A novel calibration method for the JET real-time far infrared polarimeter and integration of polarimetry-based line-integrated density measurements for machine protection of a fusion plant

In this paper, we present the work in the implementation of a new calibration for the JET real-time polarimeter based on the complex amplitude ratio technique and a new self-validation mechanism of data. This allowed easy integration of the polarimetry measurements into the JET plasma density control (gas feedback control) and as well as machine protection systems (neutral beam injection heating safety interlocks). The new addition was used successfully during 2014 JET Campaign and is envisaged that will operate routinely from 2015 campaign onwards in any plasma condition (including ITER relevant scenarios). This mode of operation elevated the importance of the polarimetry as a diagnostic tool in the view of future fusion experiments.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Achievement of Reactor-Relevant Performance in Negative Triangularity Shape in the DIII-D Tokamak

Plasma discharges with a negative triangularity (δ=−0.4) shape have been created in the DIII-D tokamak with a significant normalized beta (βN=2.7) and confinement characteristic of the high confinement mode (H98y2=1.2) despite the absence of an edge pressure pedestal and no edge localized modes (ELMs). These inner-wall-limited plasmas have a similar global performance as a positive triangularity (δ=+0.4) ELMing H-mode discharge with the same plasma current, elongation and cross sectional area. For cases both of dominant electron cyclotron heating with Te/Ti>1 and dominant neutral beam injection heating with Te/Ti=1, turbulent fluctuations over radii 0.5<ρ<0.9 were reduced by 10–50% in the negative triangularity shape compared to the matching positive triangularity shape, depending on the radius and conditions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Anomalous Fast-Ion Transport During Neutral Beam Injection on EAST

The fast-ion transport during neutral beam injection on the Experimental Advanced Superconducting Tokamak (EAST) is studied. Based on the NUBEAM and TRANSP codes, it is found that fast-ion transport is anomalous when the minimum safety factor (q{sub min}) is about 2, while it is neoclassical when q{sub min} is around 1. Neutral beam injection heating efficiency, plasma stored energy, and the total heating power are reduced when the fast ion transport is anomalous. The Alfvén continuum spectrum and the mode structures of toroidal Alfvén eigenmodes (AEs) are also calculated for comparison between neoclassical fast-ion transport and anomalous fast-ion transport. High-q{sub min} discharge with anomalous fast-ion transport has more AE activity than that of lower q{sub min} discharge with neoclassical fast-ion transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Investigation of the plasma radiation power in the Globus-M tokamak by means of SPD silicon photodiodes

Radiation losses from the plasma of the Globus-M tokamak are studied by means of SPD silicon photodiodes developed at the Ioffe Institute, Russian Academy of Sciences. The results from measurements of radiation losses in regimes with ohmic and neutral beam injection heating of plasmas with different isotope compositions are presented. The dependence of the radiation loss power on the plasma current and plasma–wall distance is investigated. The radiation power in different spectral ranges is analyzed by means of an SPD spectrometric module. Results of measurements of radiation losses before and after tokamak vessel boronization are presented. The time evolution of the sensitivity of the SPD photodiode during its two-year exploitation in Globus-M is analyzed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Benchmark of numerical tools simulating beam propagation and secondary particles in ITER NBI

Injection of high energy beams of neutral particles is a method for plasma heating in fusion devices. The ITER injector, and its prototype MITICA (Megavolt ITER Injector and Concept Advancement), are large extrapolations from existing devices: therefore numerical modeling is needed to set thermo-mechanical requirements for all beam-facing components. As the power and charge deposition originates from several sources (primary beam, co-accelerated electrons, and secondary production by beam-gas, beam-surface, and electron-surface interaction), the beam propagation along the beam line is simulated by comprehensive 3D models. This paper presents a comparative study between two codes: BTR has been used for several years in the design of the ITER HNB/DNB components; SAMANTHA code was independently developed and includes additional phenomena, such as secondary particles generated by collision of beam particles with the background gas. The code comparison is valuable in the perspective of the upcoming experimental operations, in order to prepare a reliable numerical support to the interpretation of experimental measurements in the beam test facilities. The power density map calculated on the Electrostatic Residual Ion Dump (ERID) is the chosen benchmark, as it depends on the electric and magnetic fields as well as on the evolution of the beam species via interaction with the gas. Finally the paper shows additional results provided by SAMANTHA, like the secondary electrons produced by volume processes accelerated by the ERID fringe-field towards the Cryopumps.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Achievement of reactor-relevant high normalized beta and high confinement with L-mode edge in negative triangularity shape in DIII-D

In this paper, plasma discharges with negative triangularity (δ =-0.4) shape have been created in DIII-D with significant normalized beta (ß N = 2.7) and confinement characteristic of the “high confinement” mode (H 98y2 = 1.2) despite the absence of an edge pressure pedestal and no edge localized modes (ELMs). These inner-walllimited plasmas have similar global performance as a positive triangularity (δ = +0.4) ELMing H-mode discharge with the same plasma current, elongation and cross-sectional area. For cases both of dominant electron cyclotron heating with T e / T i > 1 and dominant neutral beam injection heating with T e / T i = 1, turbulent fluctuations over radii 0.5 < ρ < 0.9 were reduced in negative triangularity over the matching positive triangularity by 10-50% depending on radius and conditions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

A wire calorimeter for the SPIDER beam: Experimental tests and feasibility study

To study and optimize negative ion production and acceleration, in view of the use of neutral beam injectors in the ITER project, the SPIDER test facility (particle energy 100keV; beam current 50A, distributed over 1280 beamlets) is under construction in Padova, with the aim of testing beam characteristics and to verify the source proper operation, by means of several diagnostic systems. An array of tungsten wires, directly exposed to the beam and consequently heated to high temperature, is used in similar experiments at IPP-Garching to study the beam optics, which is one of the most important issues, in a qualitative way. The present contribution gives a description of an experimental investigation of the behavior of tungsten wires under high heat loads in vacuum. Samples of tungsten wires are heated by electrical currents and the emitted light is measured by a camera in the 400-1100nm wavelength range, which is proposed as a calibration tool. Simultaneously, the voltage applied to the wire is measured to study the dependency of emissivity on temperature. The feasibility study of a wire calorimeter for SPIDER is finally proposed; to this purpose, the expected behaviour of tungsten with the two-dimensional beam profile in SPIDER is numerically addressed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Advanced ion beam calorimetry for the test facility ELISE

The negative ion source test facility ELISE (Extraction from a Large Ion Source Experiment) is in operation since beginning of 2013 at the Max-Planck-Institut für Plasmaphysik (IPP) in Garching bei München. The large radio frequency driven ion source of ELISE is about 1×1 m{sup 2} in size (1/2 the ITER source) and can produce a plasma for up to 1 h. Negative ions can be extracted and accelerated by an ITER-like extraction system made of 3 grids with an area of 0.1 m{sup 2}, for 10 s every 3 minutes. A total accelerating voltage of up to 60 kV is available, i.e. a maximum ion beam power of about 1.2 MW can be produced. ELISE is equipped with several beam diagnostic tools for the evaluation of the beam characteristics. In order to evaluate the beam properties with a high level of detail, a sophisticated diagnostic calorimeter has been installed in the test facility at the end of 2013, starting operation in January 2014. The diagnostic calorimeter is split into 4 copper plates with separate water calorimetry for each of the plates. Each calorimeter plate is made of 15×15 copper blocks, which act as many separate inertial calorimeters and are attached to a copper plate with an embedded cooling circuit. The block geometry and the connection with the cooling plate are optimized to accurately measure the time-averaged power of the 10 s ion beam. The surface of the blocks is covered with a black coating that allows infrared (IR) thermography which provides a 2D profile of the beam power density. In order to calibrate the IR thermography, 48 thermocouples are installed in as many blocks, arranged in two vertical and two horizontal rows. The paper describes the beam calorimetry in ELISE, including the methods used for the IR thermography, the water calorimetry and the analytical methods for beam profile evaluation. It is shown how the maximum beam inhomogeneity amounts to 13% in average. The beam divergence derived by IR thermography ranges between 1° and 4° and correlates relatively well with the values measured by beam emission spectroscopy with up to 30% difference. It has also been observed that the beam inhomogeneity decreases with lower beam divergence.

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

Powerloads on the front end components and the duct of the heating and diagnostic neutral beam lines at ITER

The heating and current drive beam lines (HNB) at ITER are expected to deliver ∼16.7 MW power per beam line for H beams at 870 keV and D beams at 1 MeV during the H-He and the DD/DT phases of ITER operation respectively. On the other hand the diagnostic neutral beam (DNB) line shall deliver ∼2 MW power for H beams at 100 keV during both the phases. The path lengths over which the beams from the HNB and DNB beam lines need to be transported are 25.6 m and 20.7 m respectively. The transport of the beams over these path lengths results in beam losses, mainly by the direct interception of the beam with the beam line components and reionisation. The lost power is deposited on the surfaces of the various components of the beam line. In order to ensure the survival of these components over the operational life time of ITER, it is important to determine to the best possible extent the operational power loads and power densities on the various surfaces which are impacted by the beam in one way or the other during its transport. The main factors contributing to these are the divergence of the beamlets and the halo fraction in the beam, the beam aiming, the horizontal and vertical misalignment of the beam, and the gas profile along the beam path, which determines the re-ionisation loss, and the re-ionisation cross sections. The estimations have been made using a combination of the modified version of the Monte Carlo Gas Flow code (MCGF) and the BTR code. The MCGF is used to determine the gas profile in the beam line and takes into account the active gas feed into the ion source and neutraliser, the HNB-DNB cross over, the gas entering the beamline from the ITER machine, the additional gas atoms generated in the beam line due to impacting ions and the pumping speed of the cryopumps. The BTR code has been used to obtain the power loads and the power densities on the various surfaces of the front end components and the duct modules for different scenarios of ITER operation. The gas profile and the magnetic field distribution for each scenario has been considered in these evaluations. The worst case power loads and power densities for each surface have been used to study their thermo-mechanical behaviour and manufacturing feasibility. The details of these calculations and results obtained are presented and discussed.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

NIO1 diagnostics

The radio frequency ion source NIO1, jointly developed by Consorzio RFX and INFN-LNL, will generate a 60kV-135mA hydrogen negative ion beam, composed of 9 beamlets over an area of about 40 × 40 mm{sup 2}. This experiment will operate in continuous mode and in conditions similar to those foreseen for the larger ion sources of the Neutral Beam Injectors for ITER. The modular design of NIO1 is convenient to address the several still open important issues related to beam extraction, optics, and performance optimization. To this purpose a set of diagnostics is being implemented. Electric and water cooling plant related measurements will allow monitoring current, pressure, flow, and temperature. The plasma in the source will be characterized by emission spectroscopy, cavity ring-down and laser absorption spectroscopy. The accelerated beam will be analyzed with a fast emittance scanner, its intensity profile and divergence with beam emission spectroscopy and visible tomography. The power distribution of the beam on the calorimeter will be monitored by thermocouples and by an infrared camera. This contribution presents the implementation and initial operation of some of these diagnostics in the commissioning phase of the experiment, in particular the cooling water calorimetry and emission spectroscopy.

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

Stabilization of electron-scale turbulence by electron density gradient in national spherical torus experiment

Theory and experiments have shown that electron temperature gradient (ETG) turbulence on the electron gyro-scale, k{sub ⊥}ρ{sub e} ≲ 1, can be responsible for anomalous electron thermal transport in NSTX. Electron scale (high-k) turbulence is diagnosed in NSTX with a high-k microwave scattering system [D. R. Smith et al., Rev. Sci. Instrum. 79, 123501 (2008)]. Here we report on stabilization effects of the electron density gradient on electron-scale density fluctuations in a set of neutral beam injection heated H-mode plasmas. We found that the absence of high-k density fluctuations from measurements is correlated with large equilibrium density gradient, which is shown to be consistent with linear stabilization of ETG modes due to the density gradient using the analytical ETG linear threshold in F. Jenko et al. [Phys. Plasmas 8, 4096 (2001)] and linear gyrokinetic simulations with GS2 [M. Kotschenreuther et al., Comput. Phys. Commun. 88, 128 (1995)]. We also found that the observed power of electron-scale turbulence (when it exists) is anti-correlated with the equilibrium density gradient, suggesting density gradient as a nonlinear stabilizing mechanism. Higher density gradients give rise to lower values of the plasma frame frequency, calculated based on the Doppler shift of the measured density fluctuations. Linear gyrokinetic simulations show that higher values of the electron density gradient reduce the value of the real frequency, in agreement with experimental observation. Nonlinear electron-scale gyrokinetic simulations show that high electron density gradient reduces electron heat flux and stiffness, and increases the ETG nonlinear threshold, consistent with experimental observations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗