A user configurable data acquisition and signal processing system for high-rate,high channel count applications

Real-time signal processing in plasma fusion experiments is required for control and for data reduction as plasma pulse times grow longer. The development time and cost for these high-rate, multichannel signal processing systems can be significant. This paper proposes a new digital signal processing (DSP) platform for the data acquisition system that will allow users to easily customize real-time signal processing systems to meet their individual requirements.The D-TACQ reconfigurable user in-line DSP (DRUID) system carries out the signal processing tasks in hardware co-processors (CPs) implemented in an FPGA, with an embedded microprocessor (μP) for control. In the fully developed platform, users will be able to choose co-processors from a library and configure programmable parameters through the μP to meet their requirements.The DRUID system is implemented on a Spartan 6 FPGA, on the new rear transition module (RTM-T), a field upgrade to existing D-TACQ digitizers.As proof of concept, a multiply-accumulate (MAC) co-processor has been developed, which can be configured as a digital chopper-integrator for long pulse magnetic fusion devices. The DRUID platform allows users to set options for the integrator, such as the number of masking samples. Results from the digital integrator are presented for a data acquisition system with 96 channels simultaneously acquiring data at 500 kSamples/s per channel.     

Overview of the COMPASS CODAC system

This paper presents an overview of the Control, Data Acquisition, and Communication system (CODAC) at the COMPASS tokamak: the hardware set-up, software implementation, and communication tools are described.The diagnostics and the data acquisition are tailored for high spatial and temporal resolution required by the COMPASS physics programme, which aims namely at studies of the plasma edge, pedestal, and Scrape-off-Layer (SOL). Studies of instabilities and turbulence are also an integral part of the programme. Therefore, the data acquisition consists of more than 1000 channels, sampled at rates from 500 kS/s up to 2 GS/s.Presently, the feedback system controls the plasma position and shape, plasma current, and density and it includes 32 analogue input channels as well as 1 digital input/output channel and 8 analogue outputs. The feedback control runs within the Multi-threaded Application Real-Time executor (MARTe) framework with two threads, a 500 μs cycle to control slow systems and a 50 μs cycle to control the fast feedback power supplies for plasma position control.In this paper, special attention is paid to the links between the systems, to the hardware and software connections, and to the communication. The hardware part is described, the software framework is addressed, and the particular implementation – the dedicated software modules, communication protocols, and links to the database are described.     

ISTTOK real-time architecture

The ISTTOK tokamak was upgraded with a plasma control system based on the Advanced Telecommunications Computing Architecture (ATCA) standard. This control system was designed to improve the discharge stability and to extend the operational space to the alternate plasma current (AC) discharges as part of the ISTTOK scientific program. In order to accomplish these objectives all ISTTOK diagnostics and actuators relevant for real-time operation were integrated in the control system.The control system was programmed in C++ over the Multi-threaded Application Real-Time executor (MARTe) which provides, among other features, a real-time scheduler, an interrupt handler, an intercommunications interface between code blocks and a clearly bounded interface with the external devices. As a complement to the MARTe framework, the BaseLib2 library provides the foundations for the data, code introspection and also a Hypertext Transfer Protocol (HTTP) server service.Taking advantage of the modular nature of MARTe, the algorithms of each diagnostic data processing, discharge timing, context switch, control and actuators output reference generation, run on well-defined blocks of code named Generic Application Module (GAM). This approach allows reusability of the code, simplified simulation, replacement or editing without changing the remaining GAMs.The ISTTOK control system GAMs run sequentially each 100 μs cycle on an Intel^® Q8200 4-core processor running at 2.33 GHz located in the ATCA crate. Two boards (inside the ATCA crate) with 32 analog-to-digital converters (ADCs) were used for acquiring the diagnostics data. Each ADC operates at 2 Msample/s but (for real-time operation) the acquired data is decimated in real-time on the board's Field-programmable gate array (FPGA) to a frequency defined by the control cycle time.This paper presents the ISTTOK real-time architecture and the human–machine Interface (HMI) for simplified AC discharge programming.     

The effect of displacement damage on deuterium retention in ITER-grade tungsten exposed to low-energy,high-flux pure and helium-seeded deuterium plasmas

Samples prepared from polycrystalline ITER-grade tungsten were damaged by irradiation with 20 MeV W ions at room temperature to a fluence of 1.4 × 10¹⁸ W/m². Due to the irradiation, displacement damage peaked near the end-of-range, 1.35 μm beneath the surface, at 0.89 displacements per atom. The damaged as well as undamaged W samples were then exposed to low-energy, high-flux (10²² D/m² s) pure D and helium-seeded D plasmas to an ion fluence of 3 × 10²⁶ D/m² at various temperatures. Trapping of deuterium was examined by the D(³He,p)⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV allowing determination of the D concentration at depths up to 6 μm. It has been found that (i) addition of 10% helium ions into the D plasma at exposure temperatures of 440–650 K significantly reduces the D concentration at depths of 0.5–6 μm compared to that for the pure plasma exposure; (ii) generation of the W-ion-induced displacement damage significantly increases the D concentration at depths up to 2 μm (i.e., in the damage zone) under subsequent exposures to both pure D and D–He plasmas.     

Biological studies using mammalian cell lines and the current status of the microbeam irradiation system,SPICE

The development of SPICE (single-particle irradiation system to cell), a microbeam irradiation system, has been completed at the National Institute of Radiological Sciences (NIRS). The beam size has been improved to approximately 5 μm in diameter, and the cell targeting system can irradiate up to 400–500 cells per minute. Two cell dishes have been specially designed: one a Si₃N₄ plate (2.5 mm × 2.5 mm area with 1 μm thickness) supported by a 7.5 mm × 7.5 mm frame of 200 μm thickness, and the other a Mylar film stretched by pressing with a metal ring. Both dish types may be placed on a voice coil stage equipped on the cell targeting system, which includes a fluorescent microscope and a CCD camera for capturing cell images. This microscope system captures images of dyed cell nuclei, computes the location coordinates of individual cells, and synchronizes this with the voice coil motor stage and single-particle irradiation system consisting of a scintillation counter and a beam deflector. Irradiation of selected cells with a programmable number of protons is now automatable. We employed the simultaneous detection method for visualizing the position of mammalian cells and proton traversal through CR-39 to determine whether the targeted cells are actually irradiated. An immuno-assay was also performed against γ-H2AX, to confirm the induction of DNA double-strand breaks in the target cells.     

Rearranging a nanoprobe: Line foci,grid shadow patterns and performance tests

After a major modification of the target chamber at the Leipzig high energy ion nanoprobe the probe forming lens system, consisting of two separated quadrupole doublets, had been carefully realigned. This was done by adjusting the line foci position of each individual quadrupole on the centre position defined by the unfocused beam. Using a high magnification microscope the alignment process is very effective and precise. The lens system could be precisely realigned except an intrinsic rotational misalignment which is essentially reduced by a correction lens.Grid shadow patterns have been taken and analysed in order to assess the characteristics of the system. The dominant aberrations are spherical with an additional parasitic octupole.The grid shadow method is also very useful to determine the best position of the aperture diaphragms which minimizes the influence of the aberrations onto the beam spot size.The rearrangement allowed larger aperture diaphragms for higher beam currents at a moderate increase in beam spot sizes. Performance tests yielded proton microbeam currents and half-widths of 4.5 nA at 1.5 μm, 8.3 nA at 1.5 μm and 17.2 nA at 2 μm. For high resolution work the expected beam spots around 0.3 μm at 100 pA were not achieved. The reason is very likely interference on the beam scanner, correlated in x- and y-direction, which results from the insufficiently rectified power supply voltage of the transconductance amplifier.     

Commissioning and initial operation of KSTAR superconducting tokamak

The commissioning and the initial operation for the first plasma in the KSTAR device have been accomplished successfully without any severe failure preventing the device operation and plasma experiments. The commissioning is classified into four steps: vacuum commissioning, cryogenic cool-down commissioning, magnet system commissioning, and plasma discharge.Vacuum commissioning commenced after completion of the tokamak and basic ancillary systems construction. Base pressure of the vacuum vessel was about 3 × 10^?6 Pa and that of the cryostat about 2.7 × 10^?4 Pa, and both levels meet the KSTAR requirements to start the cool-down operation. All the SC magnets were cooled down by a 9 kW rated cryogenic helium facility and reached the base temperature of 4.5 K in a month. The performance test of the superconducting magnet showed that the joint resistances were below 3 nΩ and the resistance to ground after cool-down was over 1 GΩ. An ac loss test of each PF coil made by applying a dc biased sinusoidal current showed that the coupling loss was within the KSTAR requirement with the coupling loss time constant less than 35 ms for both Nb₃Sn and NbTi magnets. All the superconducting magnets operated in stable without quench for long-time dc operation and with synchronized pulse operation by the plasma control system (PCS). By using an 84 GHz ECH system, second harmonic ECH assisted plasma discharges were produced successfully with loop voltage of less than 3 V. By the real-time feedback control, operation of 100 kA plasma current with pulse length up to 865 ms was achieved, which also meet the first plasma target of 100 kA and 100 ms. The KSTAR device will be operated to meet the missions of steady-state and high-beta achievement by system upgrades and collaborative researches.     

Concept design on RH maintenance of CFETR Tokamak reactor

CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.     

ASDEX Upgrade enhancements in view of ITER application

The aim of the ASDEX Upgrade (AUG) programme is to support the design, prepare the physics base and develop regimes beyond the baseline of ITER and for DEMO. Its ITER-like geometry, poloidal field system, versatile heating system and power fluxes make AUG particularly suited.After the transition to fully tungsten coated plasma facing components AUG could be operated without prior boronizations and a low permanent deuterium retention was found qualifying W as wall material. ITER-like baseline H-modes (H₉₈ ～ 1, β_N ～ 2) were routinely achieved up to 1.2 MA plasma currents. W concentrations could be kept at an acceptable level of <5 × 10^?5 by central wave heating (enhancing impurity outward transport) and ELM pacing with gas puffing. The compatibility of high performance improved H-modes, the ITER hybrid scenario, with an un-boronized W wall was demonstrated achieving H₉₈ ～ 1.1 and β_N up to 2.6 at modest triangularities δ ≤ 0.3. This performance is reached despite the gas puffing needed for W influx control. Increasing δ to 0.35 allowed at even higher puff rates still a H₉₈ ～ 1.1.Reliable plasma operation in support of ITER comprised the demonstration of ECRF assisted low voltage plasma start-up and current rise at toroidal electric fields below 0.3 V/m resulting in a ITER compatible range of plasma internal inductance of 0.71–0.97. Disruption mitigation is feasible using strong gas puffs, and the achieved electron densities approach values needed for runaway suppression.Present hardware extensions in support of ITER include the upgrading of ECRH by a 4 MW/10 s system with large deposition variability (tuneable frequency between 105 and 140 GHz, real-time steerable mirrors) for central heating and MHD mode control. A powerful system of 24 in-vessel coils produces error fields up to toroidal mode number n = 4 for ELM suppression and mode rotation control. In connection with a close conducting wall they will open up the road for RWM stabilization in advanced scenarios. For those we are considering LHCD for current drive and profile control with up to 500 kA driven current. The tungsten sources are dominated by sputtering from intrinsic light impurities, and the W influx from the outboard limiters are the main source for the core plasma. ICRH induced electric fields accelerate light impurities, restricting the use of ICRH to just after boronization. 4-strap antennas imbedded in extended wall structures might solve this problem. Finally, doubling the plasma volume with plasma currents above 2 MA in AUG could be the solution for a needed ITER satellite.     

Preparation of tungsten coatings on graphite by electro-deposition via Na2WO4–WO3 molten salt system

Tungsten coating on graphite substrate is one of the most promising candidate materials as the ITER plasma facing components. In this paper, tungsten coatings on graphite substrates were fabricated by electro-deposition from Na₂WO₄–WO₃ molten salt system at 1173 K in atmosphere. Tungsten coatings with no impurities were successfully deposited on graphite substrates under various pulsed current densities in an hour. By increasing the current density from 60 mA cm⁻² to 120 mA cm⁻² an increase of the average size of tungsten grains, the thickness and the hardness of tungsten coatings occurs. The average size of tungsten grains can reach 7.13 μm, the thickness of tungsten coating was in the range of 28.8–51 μm, and the hardness of coating was higher than 400 HV. No cracks or voids were observed between tungsten coating and graphite substrate. The oxygen content of tungsten coating is about 0.022 wt%.     

Carbon film growth and hydrogenic retention of tungsten exposed to carbon-seeded high density deuterium plasmas

Tungsten (W) targets have been exposed to high density (n_e ? 4 × 10¹⁹ m^?3), low temperature (T_e ? 3 eV) CH₄-seeded deuterium (D) plasma in Pilot-PSI. The surface temperature of the target was ～1220 K at the center and decreased radially to ～650 K at the edges. Carbon film growth was found to only occur in regions where there was a clear CII emission line, corresponding to regions in the plasma with T_e ? 2 eV. The maximum film thickness was ～2.1 μm after a plasma exposure time of 120 s. ³He nuclear reaction (NRA) analysis and thermal desorption spectroscopy (TDS) determine that the presence of a thin carbon film dominates the hydrogenic retention properties of the W substrate. Thermal desorption spectroscopy analysis shows retention increasing roughly linearly with incident plasma fluence. NRA measures a C/D ratio of ～0.002 in these films deposited at high surface temperatures.     

Improvement of initial vacuum condition along 2008–2010 KSTAR campaign by vessel baking

Korea Superconducting Tokamak Advanced Research (KSTAR) is upgraded for its KSTAR 3rd campaign for new target mission to produce the D-shaped plasma with a target plasma current of 500 kA and/or pulse length of 5 s. New Plasma Facing Components (PFCs) are installed which leads to the increase of the surface area of the vessel by a factor of about 5. The vacuum conditioning such as the vessel baking has been performed in order to remove various kinds of impurities including H₂O, carbon and oxygen for the plasma. The total outgassing rate in the KSTAR 1st campaign was measured as 1.5 × 10^?4 mbar ? s^?1 which is increased by a factor of 3 (6.49 × 10^?4 mbar ? s^?1) in the KSTAR 3rd campaign. Nevertheless, the outgassing rates per unit area have been decreased from 9.31 × 10^?5 mbar ? m^?2 s^?1 to 1.22 × 10^?5 mbar ? m^?2 s^?1 due to the upgrade of baking system and series of baking operation.     

The ARIES-AT advanced tokamak,Advanced technology fusion power plant

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R&D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β_N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ_x = 2.2) which is the result of a “thinner” blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β_N. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb–17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 °C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb–17Li to about 1000 °C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to an attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (4.7 ¢/kWh), which is competitive with those projected for other sources of energy.     

Porous iron pellets for AMS 14C analysis of small samples down to ultra-microscale size (10–25 μgC)

We developed the use of a porous iron pellet as a catalyst for AMS ¹⁴C analysis of small samples down to ultra-microscale size (10–25 μgC). It resulted in increased and more stable beam currents through our HVEE 4130 ¹⁴C AMS system, which depend smoothly on the sample size. We find that both the expected decrease of oxalic acid standards and increase of backgrounds with decreasing sample size, due to increasing influence of contamination, are reproducible. Using a mass-dependent background correction for dead (1.0 ± 0.4 μgC) and modern (0.25 ± 0.10 μgC) contamination, we obtain reliable results for small samples down to 10 μgC and possibly smaller. Due to our low graphitization yield for ultra-small samples (increases from 40% to 80% on average with sample size), we measured graphite standards as small as 3 μgC. The standard deviation of the corrected activity is about 5% for a 10-μgC HOxII standard.Here we report the iron pellet technique, which is new to the best of our knowledge. It is generally applicable for AMS ¹⁴C laboratories that want to measure small samples down to ultra-microscale size. As an illustrative test-case, we analyze ¹⁴C data for IAEA-C5, C7 and C8 samples with masses ranging from 15 to 300 μgC.     

Application of laser-induced breakdown spectroscopy for characterization of material deposits and tritium retention in fusion devices

Laser-induced breakdown spectroscopy (LIBS) is discussed as a possible method to characterize the composition, tritium retention and amount of material deposits on the first wall of fusion devices. The principle of the technique is the ablation of the co-deposited layer by a laser pulse with P (power density) ≥ 0.5 GW/cm² and the spectroscopic analysis of the light emitted by the laser induced plasma. The typical spatial extension of the laser plasma plume is in the order of 1 cm with typical plasma parameters of n_e ≈ 3 × 10²² m^?3 and T_e ≈ 1–2 eV averaged over the plasma lifetime which is below 1 μs. In this study “ITER-Like” mixed deposits with a thickness of about 2 μm and consisting of a mixture of W/Al/C and D on bulk tungsten substrates have been analyzed by LIBS to measure the composition and hydrogen isotopes content at different laser energies, ranging from about 2 J/cm² (0.3 GW/cm²) to about 17 J/cm² (2.4 GW/cm²) for 7 ns laser pulses. It is found that the laser energies above about 7 J/cm² (1 GW/cm²) are needed to achieve the full removal of the deposit layer and identify a clear interface between the deposit and the bulk tungsten substrate by applying 15–20 laser pulses while hydrogen isotopes decrease strongly after the first laser pulse. Under these conditions, the evolution of the spectral line intensities of W/Al/C/hydrogen can be used to evaluate the layer composition.     

Development of high voltage power supply for the KSTAR 170 GHz ECH and CD system

A 3.6 MW (66 kV/55 A) DC power supply system was developed for the 170 GHz EC H&CD system in KSTAR. The power supply system consists of a cathode power supply (CPS), an anode power supply (APS) and a body power supply (BPS). The cathode power supply is capable of supplying a maximum voltage of ?66 kV and a current of 55 A to the cathode with respect to the collector using pulse step modulation (PSM). The high voltage switching system for the cathode is made by a fast MOS-FET solid-state switch which can turn off the high voltage to the cathode within 3 μs in the occurrence of gyrotron faults. The APS is a voltage divider system consisting of a fixed resistor and zener diode units with the capability of 60 kV stand-off voltage. The anode voltage with respect to the cathode is controlled in a range of 0–60 kV by turning the MOS-FET switches connected in parallel to each zener diode on and off. For high frequency current modulation of the gyrotron, the parallel discharge switch is introduced between the cathode and anode in order to clamp the charged voltage in the stray capacitance. The BPS is a DC power supply with the capability of 50 kV/160 mA. The nominal operation parameter of BPS was 23 kV and 10 mA, respectively, and the voltage output is regulated with a stability of 0.025% of the rated voltage. The series MOS-FET solid-state switch is used for on/off modulation in the body voltage sychronizing with anode voltage. The parallel discharge switch is also introduced between the body and collector for high frequency RF modulation. This paper describes the key features of the high voltage power supply system of the KSTAR 170 GHz gyrotron as well as the test results of the power supply.     

Design of FPGA based high-speed data acquisition and real-time data processing system on J-TEXT tokamak

Tokamak experiment requires high-speed data acquisition and processing systems. In traditional data acquisition system, the sampling rate, channel numbers and processing speed are limited by bus throughput and CPU speed. This paper presents a data acquisition and processing system based on FPGA. The data can be processed in real-time before it is passed to the CPU. It provides processing ability for more channels with higher sampling rates than the traditional data acquisition system while ensuring deterministic real-time performance. A working prototype is developed for the newly built polarimeter–interferometer diagnostic system on the Joint Texas Experimental Tokamak (J-TEXT). It provides 16 channels with 120 MHz maximum sampling rate and 16 bit resolution. The onboard FPGA is able to calculate the plasma electron density and Faraday rotation angel. A RAID 5 storage device is adopted providing 700 MB/s read–write speed to buffer the data to the hard disk continuously for better performance.     

FIRE,a novel concept of massive gas injection for disruption mitigation in ITER: Validation on Tore Supra

Disruptions in large size tokamak like ITER must be mitigated to reduce detrimental effects on the device. Massive impurity injection prior to disruption is a promising mitigation technique. Many injector designs have already been tested on nowadays tokamaks. A novel concept of injector has been designed and tested on Tore Supra. It is based on the use of high pressure cartridges sealed with a bursting disc. For firing, an electric arc is generated inside the cartridge close to the rupture disc. This initiates a shock wave which breaks the disc. Tests show that the opening is achieved in less than 400 μs. A detailed modeling of the fast injector operation is presented. One result is that the gas is expelled in about 2 ms and an outflow rate as large as 4.2 × 10²⁵ atoms/s can be achieved at the nozzle exit (1 cm diameter). Massive neon gas injections have be performed on stable plasma to validate the concept. The gas penetration into the plasma prior to disruption is followed by a fast camera. Comparison with a ferromagnetic-valve based injector was also carried out. It is found that the velocity of the cold front penetration into the hot plasma is larger using this new injector.     

Experimental set-up for time resolved small angle X-ray scattering studies of nanoparticles formation using a free-jet micromixer

Recently, we have designed, fabricated and tested a free-jet micromixer for time resolved small angle X-ray scattering (SAXS) studies of nanoparticles formation in the <100 μs time range. The microjet has a diameter of 25 μm and a time of first accessible measurement of 75 μs has been obtained. This result can still be improved. In this communication, we present a method to estimate whether a given chemical or biological reaction can be investigated with the micromixer, and to optimize the beam size for the measurement at the chosen SAXS beamline. Moreover, we describe a system based on stereoscopic imaging which allows the alignment of the jet with the X-ray beam with a precision of 20 μm. The proposed experimental procedures have been successfully employed to observe the formation of calcium carbonate (CaCO₃) nanoparticles from the reaction of sodium carbonate (Na₂CO₃) and calcium chloride (CaCl₂). The induction time has been estimated in the order of 200 μs and the determined radius of the particles is about 14 nm.