Diagnostic Neutral Beam Research Articles

Spectroscopic determination of the composition of a 50 kV hydrogen diagnostic neutral beam.

A grating spectrometer with an electron multiplying charge-coupled device camera is used to diagnose a 50 kV, 5 A, 20 ms hydrogen diagnostic neutral beam. The ion source density is determined from Stark broadened Hβ emission and the spectrum of Doppler-shifted Hα emission is used to quantify the fraction of ions at full, half, and one-third beam energy under a variety of operating conditions including fueling gas pressure and arc discharge current. Beam current is optimized at low-density conditions in the ion source while the energy fractions are found to be steady over most operating conditions.

Sensitivity of MSE measurements on the beam atomic level population.

The effect of variation in atomic level population of a neutral beam on the Motional Stark Effect (MSE) measurements is investigated in the low density plasmas of HSX stellarator. A 30 KeV, 4 A, 3 ms hydrogen diagnostic neutral beam is injected into HSX plasmas of line averaged electron density ranging from 2 to 4 ⋅ 1018 m-3 at a magnetic field of 1 T. For this density range, the excited level population of the hydrogen neutral beam is expected to undergo variations. Doppler shifted and Stark split Hα and Hβ emissions from the beam are simultaneously measured using two cross-calibrated spectrometers. The emission spectrum is simulated and fit to the experimental measurements and the deviation from a statistically populated beam is investigated.

Measurement of diagnostic neutral beam parameters on J-TEXT.

A Doppler frequency shift spectrum (DFSS) system composed of two spectrometers has been developed for the joint Texas experimental tokamak to measure diagnostic neutral beam parameters including the beam energy fractions, intensity distributions, and divergences. The beam energy fractions are derived from measurements of H-alpha (Hα) emission using collisional excitation cross sections. The beam intensity distributions are obtained using an 11-channel measurement with a reconstruction technique. The beam divergences are obtained from spectrum broadening and geometric calculations. The results of preliminary investigations indicate that the DFSS system works well and can be used to obtain all of these parameters simultaneously. According to the preliminary experiment, the one-third energy fraction has the largest proportion (about 45%) of the beam energy and the full energy fraction is about 10%. The beam diameter is about 8.1 cm at a distance of 2.04 m from the accelerator. The beam divergence angle is about 3.3°. The current beam parameters are insufficient for charge-exchange measurements.

Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines.The experimental phases involve application of HV power supplies (100KV) and High RF power (∼800KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations.To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as prescribed by the ITER PCDH guidelines.

Charge exchange recombination spectroscopy on the T-10 tokamak.

The charge exchange recombination spectroscopy (CXRS) diagnostics on the T-10 tokamak is described. The system is based on a diagnostic neutral beam and includes three high etendue spectrometers designed for the ITER edge CXRS system. A combined two-channel spectrometer is developed for simultaneous measurements of two beam-induced spectral lines using the same lines of sight. A basic element of the combined spectrometer is a transmitting holographic grating designed for the narrow spectral region 5291 ± 100 Å. The whole CXRS system provides simultaneous measurements of two CXRS impurity spectra and Hα beam line. Ion temperature measurements are routinely provided using the C(6+) CXRS spectral line 5291 Å. Simultaneous measurements of carbon densities and one more impurity (oxygen, helium, lithium etc.) are carried out. Two light collecting systems with 9 lines of sight in each system are used in the diagnostics. Spatial resolution is up to 2.5 cm and temporal resolution of 1 ms is defined by the diagnostic neutral beam diameter and pulse duration, respectively. Experimental results are shown to demonstrate a wide range of the CXRS diagnostic capabilities on T-10 for investigation of impurity transport processes in tokamak plasma. Developed diagnostics provides necessary experimental data for studying of plasma electric fields, heat and particle transport processes, and for investigation of geodesic acoustic modes.

Specific features of measuring the isotopic composition of hydrogen ions in ITER plasma by using neutral particle diagnostics under neutral beam injection conditions

Results of numerical simulation of signals from neutral particle analyzers under injection of the heating and diagnostic neutral beams in different operating modes of the ITER tokamak are presented. The distribution functions of fast ions in plasma are simulated, and the corresponding neutral particle fluxes escaping from the plasma along the line of sight of the analyzers are calculated. It is shown that the injection of heating deuterium (D0) beams results in the appearance of an intense background signal hampering measurements of the ratio between the densities of deuterium and tritium fuel ions in plasma in the thermal energy range. The injection of a diagnostic hydrogen (H0) beam does not affect measurements owing to the high mass resolution of the analyzers.

Overview of ion source characterization diagnostics in INTF.

INdian Test Facility (INTF) is envisaged to characterize ITER diagnostic neutral beam system and to establish the functionality of its eight inductively coupled RF plasma driver based negative hydrogen ion source and its beamline components. The beam quality mainly depends on the ion source performance and therefore, its diagnostics plays an important role for its safe and optimized operation. A number of diagnostics are planned in INTF to characterize the ion source performance. Negative ions and its cesium contents in the source will be monitored by optical emission spectroscopy (OES) and cavity ring down spectroscopy. Plasma near the extraction region will be studied using standard electrostatic probes. The beam divergence and negative ion stripping losses are planned to be measured using Doppler shift spectroscopy. During initial phase of ion beam characterization, carbon fiber composite based infrared imaging diagnostics will be used. Safe operation of the beam will be ensured by using standard thermocouples and electrical voltage-current measurement sensors. A novel concept, based on plasma density dependent plasma impedance measurement using RF electrical impedance matching parameters to characterize the RF driver plasma, will be tested in INTF and will be validated with OES data. The paper will discuss about the overview of the complete INTF diagnostics including its present status of procurement, experimentation, interface with mechanical systems in INTF, and integration with INTF data acquisition and control systems.

Overview of the negative ion based neutral beam injectors for ITER.

The ITER baseline foresees 2 Heating Neutral Beams (HNB's) based on 1 MeV 40 A D(-) negative ion accelerators, each capable of delivering 16.7 MW of deuterium atoms to the DT plasma, with an optional 3rd HNB injector foreseen as a possible upgrade. In addition, a dedicated diagnostic neutral beam will be injecting ≈22 A of H(0) at 100 keV as the probe beam for charge exchange recombination spectroscopy. The integration of the injectors into the ITER plant is nearly finished necessitating only refinements. A large number of components have passed the final design stage, manufacturing has started, and the essential test beds-for the prototype route chosen-will soon be ready to start.

Spatial calibration of a tokamak neutral beam diagnostic using in situ neutral beam emission.

Neutral beam injection is used in tokamaks to heat, apply torque, drive non-inductive current, and diagnose plasmas. Neutral beam diagnostics need accurate spatial calibrations to benefit from the measurement localization provided by the neutral beam. A new technique has been developed that uses in situ measurements of neutral beam emission to determine the spatial location of the beam and the associated diagnostic views. This technique was developed to improve the charge exchange recombination (CER) diagnostic at the DIII-D tokamak and uses measurements of the Doppler shift and Stark splitting of neutral beam emission made by that diagnostic. These measurements contain information about the geometric relation between the diagnostic views and the neutral beams when they are injecting power. This information is combined with standard spatial calibration measurements to create an integrated spatial calibration that provides a more complete description of the neutral beam-CER system. The integrated spatial calibration results are very similar to the standard calibration results and derived quantities from CER measurements are unchanged within their measurement errors. The methods developed to perform the integrated spatial calibration could be useful for tokamaks with limited physical access.

Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9m with diameter of ∼4.5m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3m to 5m with diameter of ∼1.5m and one vessel at the end to house the second calorimeter.The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

Manufacturing of the full size prototype of the ion source for the ITER neutral beam injector – The SPIDER beam source

In ITER, each heating neutral beam injector (HNB) will deliver about 16.5MW heating power by accelerating a 40A deuterium negative ion beam up to the energy of 1MeV. The ions are generated inside a caesiated negative ion source, where the injected H2/D2 is ionized by a radio frequency electromagnetic field.The SPIDER test bed, currently being manufactured, is going to be the ion source test facility for the full size ion source of the HNBs and of the diagnostic neutral beam injector of ITER.The SPIDER beam source comprises an ion source with 8 radio-frequency drivers and a three-grid system, providing an overall acceleration up to energies of about 100keV [1]. SPIDER represents a substantial step forward between the half ITER size ion source, which is currently being tested at the ELISE test bed in IPP-Garching, and the negative ion sources to be used on ITER, in terms of layout, dimensions and operating parameters. The SPIDER beam source will be housed inside a vacuum vessel which will be equipped with a beam dump and a graphite diagnostic calorimeter.The manufacturing design of the main parts of the SPIDER beam source has been completed and many of the tests on the prototypes have been successfully passed. The most complex parts, from the manufacturing point of view, of the ion source and the accelerator, developed by galvanic deposition of copper are being manufactured. The manufacturing phase will be completed within 2015, when the assembly of the device will start at the PRIMA site, in Padova (I).The paper describes the status of the procurement, the adaptations operated on the design of the beam source for the fabrication, with particular emphasis to the engineering development to enable the fulfillment of the tight requirements set in the technical specifications. Moreover the tests performed on the prototypes are reported.

Some aspects of the design of the ITER NBI Active Correction and Compensation Coils

The neutral beam system for ITER consists of two heating and current drive injectors plus a diagnostic neutral beam injector. The proposed physical plant layout allows for a possible third heating injector to be installed later. For correct operation of the beam source, and to avoid deflections of the charged fraction of the beam, the magnetic field along the beam path must be very low. To minimize the stray ITER field in critical areas (ion source, acceleration grids, neutralizer, residual ion dump), a Magnetic Field Reduction System will envelop the beam vessels and the high voltage transmission lines to ion source. This whole system comprises the Passive Magnetic Shield, a set of thick steel plates, and the Active Correction and Compensation Coils, a set of coils carrying currents which depend on the tokamak stray field. This paper describes the status of the coil design, terminals and support structures, as well as a description of the calculations carried out. Most coils are suitable for removal from their final position to be replaced in case of a fault. Conclusions of the chosen design highlight the strategy for the system feasibility.

Assembly and gap management strategy for the ITER NBI vessel passive magnetic shield

The neutral beam system for ITER consists of two heating and current drive neutral ion beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. The HNB Passive Magnetic Shield (PMS) works in conjunction with the active compensation/correction coils to limit the magnetic field inside the Beam Line Vessel (BLV), Beam Source Vessel (BSV), High Voltage Bushing (HVB) and Transmission Line (TL) elbow to acceptable levels that do not interfere with the operation of the HNB components. This paper describes the current design of the PMS, having had only minor modifications since the preliminary design review (PDR) held in IO in April 2013, and the assembly strategy for the vessel PMS.

Magnetic analysis of the magnetic field reduction system of the ITER neutral beam injector

The neutral beam system for ITER consists of two heating and current drive neutral beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. For the correct operation of the beam, the ion source and the ion path until it is neutralized must operate under a very low magnetic field environment. To prevent the stray ITER field from penetrating inside those mentioned critical areas, a magnetic field reduction system (MFRS) will envelop the beam vessels and the high voltage transmission lines to ion source. This system comprises the passive magnetic shield (PMS), a box like assembly of thick low carbon steel plates, and the Active Correction and Compensation Coils (ACCC), a set of coils carrying a current which depends on the tokamak stray field. This paper describes the magnetic model and analysis results presented at the PMS and ACCC preliminary design review held in ITER organization in April 2013. The paper focuses on the magnetic model description and on the description of the analysis results. The iterative process for obtaining optimized currents in the coils is presented. The set of coils currents chosen among the many possible solutions, the magnetic field results in the interest regions and the fulfillment of the magnetic field requirements are described.

Design of Data Acquisition and Control System for Indian Test Facility of Diagnostics Neutral Beam

The Indian Test Facility (INTF) – a negative hydrogen ion based 100kV, 60A, 5Hz modulated NBI system having 3s ON/20s OFF duty cycle. Prime objective of the facility is to install a full-scale test bed for the qualification of all Diagnostic Neutral Beam (DNB) parameters, prior to installation in ITER.The automated and safe operation of the INTF will require a reliable and rugged instrumentation and control system which provide control, data acquisition (DAQ), interlock and safety functions, referred as INTF-DACS. The INTF-DACS has been decided to be design based on the ITER CODAC architecture and ITER-PCDH guidelines since the technical understanding of CODAC technology gained from this will later be helpful in development of plant system I&C for DNB.For complete operation of the INTF, approximately 900 numbers of signals are required to be superintending by the DACS. In INTF conventional control loop time required is within the range of 5–100ms and for DAQ except high-end diagnostics, required sampling rates in range of 5 sample per second (Sps) to 10kSps; to fulfill these requirements hardware components have been selected from the ITER slow and fast controller catalogs. For high-end diagnostics required sampling rates up to 100MSps normally in case of certain events, therefore event and burst based DAQ hardware has been finalized. Combined use of CODAC core software (CCS) and NI-LabVIEW has been finalized due to the fact that full required DAQ support is not available in present version of CCS. Interlock system for investment protection of facility and Safety system for occupational safety of the involved personal, are under design as per ITER-PCDH recommendations, which are based on IEC 61508 standard.This paper described the detail design of INTF DACS based on ITER-PCDH guidelines.

A Spectrally Resolved Motional Stark Effect Diagnostic for the TJ‐II Stellarator

AbstractA spectrally resolved motional stark effect (MSE) diagnostic has been implemented for the TJ‐II stellarator to quantify the magnitude and pitch of components of the magnetic field created in this magnetic confinement device. The system includes a compact diagnostic neutral beam injector (DNBI) that provides a short pulse of accelerated neutral hydrogen atoms with an e–1 beam radius of 2.1 cm to stimulate the Doppler‐displaced Balmer Hαs emissions, which are the basis for this diagnostic. Measurement of the wavelength separation of the Stark splitting of the Hα spectrum, as well as of the relative line intensities of its components, allow the local magnitude and direction of the internal magnetic field components to be measured at 10 positions across the plasma. The use of a DNBI extends such measurements to the electron cyclotron resonance (ECR) heated phases of plasmas while also overcoming the need for the complicated inversion techniques that are required when such measurements are performed with a heating neutral beam injector (NBI). Moreover, the use of the shot‐to‐shot technique with reproducible discharges further simplifies fits to the MSE spectra as nearby impurity spectral emission lines can be eliminated or significantly reduced. After outlining the principles of this technique and the diagnostic set‐up, magnetic field measurements made during ECR or NBI heating phases are reported for a range of magnetic configurations and are compared with vacuum magnetic field estimates in order to evaluate the capabilities and limitations of this diagnostic for the TJ‐II. (© 2015 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Design of charge exchange recombination spectroscopy for the joint Texas experimental tokamak.

The old diagnostic neutral beam injector first operated at the University of Texas at Austin is ready for rejoining the joint Texas experimental tokamak (J-TEXT). A new set of high voltage power supplies has been equipped and there is no limitation for beam modulation or beam pulse duration henceforth. Based on the spectra of fully striped impurity ions induced by the diagnostic beam the design work for toroidal charge exchange recombination spectroscopy (CXRS) system is presented. The 529 nm carbon VI (n = 8 - 7 transition) line seems to be the best choice for ion temperature and plasma rotation measurements and the considered hardware is listed. The design work of the toroidal CXRS system is guided by essential simulation of expected spectral results under the J-TEXT tokamak operation conditions.

Charge-exchange recombination spectroscopy of the plasma ion temperature at the T-10 tokamak

Charge-exchange recombination spectroscopy (CXRS) based on a diagnostic neutral beam has been developed at the T-10 tokamak. The diagnostics allows one to measure the ion temperature profile in the cross section of the plasma column. In T-10 experiments, the measurement technique was adjusted and the elements of the CXRS diagnostics for ITER were tested. The used spectroscopic equipment makes it possible to reliably determine the ion temperature from the Doppler broadening of impurity lines (helium, carbon), as well as of the spectral lines of the working gas. The profiles of the plasma ion temperature in deuterium and helium discharges were measured at different plasma currents and densities, including with the use of active Doppler measurements of lines of different elements. The validity and reliability of ion temperature measurements performed by means of the developed CXRS diagnostics are analyzed.

Assembly process of the ITER neutral beam injectors

The ITER neutral beam (NB) injectors are used for heating and diagnostics operations. There are 4 injectors in total, 3 heating neutral beam injectors (HNBs) and one diagnostic neutral beam injector (DNB). Two HNBs and the DNB will start injection into ITER during the hydrogen/helium phase of ITER operations. A third HNB is considered as an upgrade to the ITER heating systems, and the impact of the later installation and use of that injector have to be taken into account when considering the installation and assembly of the whole NB system. It is assumed that if a third HNB is to be installed, it will be installed before the nuclear phase of the ITER project.The total weight of one injector is around 1200t and it is composed of 18 main components and 36 sets of shielding plates. The overall dimensions are length 20m, height 10m and width 5m.Assembly of the first two HNBs and the DNB will start before the first plasma is produced in ITER, but as the time required to assemble one injector is estimated at around 1.5 year, the assembly will be divided into 2 steps, one prior to first plasma, and the second during the machine second assembly phase. To comply with this challenging schedule the assembly sequence has been defined to allow assembly of three first injectors in parallel. Due to the similar design between the DNB and HNBs it has been decided to use the same tools, which will be designed to accommodate the differences between the two sets of components. This reduces the global cost of the assembly and the overall assembly time for the injector system.The alignment and positioning of the injectors is a major consideration for the injector assembly as the alignment of the beamline components and the beam source are critical if good injector performance is to be achieved. The theoretical axes of the beams are defined relative to the duct liners which are installed in the NB ports. The concept adopted to achieve the required alignment accuracy is to use the main rail of the overhead crane associated with offset tooling when necessary. The overhead crane is used for the assembly of the components, and the final positioning of the beamline components and the beam source will be adjusted with respect to laser targets referring to the optimum beam axis and source position.This paper describes the installation tasks and the alignment and positioning solutions and the complexity of operations within the NB cell. Particular constraints on the HNB installation sequence due to the planned testing of the 1MV high voltage supply are also described.

Collisionality scaling of main-ion toroidal and poloidal rotation in low torque DIII-D plasmas

In tokamak plasmas with low levels of toroidal rotation, the radial electric field Er is a combination of pressure gradient and toroidal and poloidal rotation components, all having similar magnitudes. In order to assess the validity of neoclassical poloidal rotation theory for determining the poloidal rotation contribution to Er, Dα emission from neutral beam heated tokamak discharges in DIII-D (Luxon 2002 Nucl. Fusion 42 614) has been evaluated in a sequence of low torque (electron cyclotron resonance heating and balanced diagnostic neutral beam pulse) discharges to determine the local deuterium toroidal rotation velocity. By invoking the radial force balance relation the deuterium poloidal rotation can be inferred. It is found that the deuterium poloidal flow exceeds the neoclassical value in plasmas with collisionality , being more ion diamagnetic, and with a stronger dependence on collisionality than neoclassical theory predicts. At low toroidal rotation, the poloidal rotation contribution to the radial electric field and its shear is significant. The effect of anomalous levels of poloidal rotation on the radial electric field and cross-field heat transport is investigated for ITER parameters.