Negative ion beam extraction in ROBIN

The RF based single driver ?ve ion source experiment test bed ROBIN (Replica Of BATMAN like source in INDIA) has been set up at Institute for Plasma Research (IPR), India in a technical collaboration with IPP, Garching, Germany. A hydrogen plasma of density 5 × 10¹² cm^?3 is expected in driver region of ROBIN by launching 100 kW RF power into the driver by 1 MHz RF generator. The cesiated source is expected to deliver a hydrogen negative ion beam of 10 A at 35 kV with a current density of 35 mA/cm² as observed in BATMAN.In first phase operation of the ROBIN ion source, a hydrogen plasma has been successfully generated (without extraction system) by coupling 80 kW RF input power through a matching network with high power factor (cos θ > 0.8) and different plasma parameters have been measured using Langmuir probes and emission spectroscopy. The plasma density of 2.5 × 10¹¹ cm^?3 has been measured in the extraction region of ROBIN. For negative hydrogen ion beam extraction in second phase operation, extraction system has been assembled and installed with ion source on the vacuum vessel. The source shall be first operated in volume mode for negative ion beam extraction. The commissioning of the source with high voltage power supply has been initiated.     

Study and optimization of boronization in Alcator C-Mod using the Surface Science Station (S3)

A Surface Science Station (S³) on the Alcator C-Mod tokamak is used to study and optimize the location and rate of boron film deposition in situ during electron cyclotron (EC) discharge plasmas using 2.45 GHz radio-frequency (RF) heating and a mixture of helium and diborane (B₂D₆) gasses. The radial profile of boron deposition is measured with a pair of quartz microbalances (QMB) on S³, the faces of which can be rotated 360° including orientations parallel and perpendicular to the toroidal magnetic field B_T ～0.1 T. The plasma electron density is measured with a Langmuir probe, also on S³ in the vicinity of the QMBs, and typical values are ～1 × 10¹⁶ m^?3. A maximum boron deposition rate of 0.82 μg/cm²/min is obtained, which corresponds to 3.5 nm/min if the film density is that of solid boron. These deposition rates are sufficient for boron film applications between tokamak discharges. However the deposition does not peak at the EC resonance as previously assumed. Rather, deposition peaks near the upper hybrid (UH) resonance, ～5 cm outboard of the EC resonance. This has implications for RF absorption, with the RF waves being no longer damped on the electrons at the EC resonance. The previously inferred radial locations of critical erosion zones in Alcator C-Mod also need to be re-evaluated. The boron deposition profile versus major radius follows the ion flux/density profile, implying that the boron deposition is primarily ionic. The application of a vertical magnetic field (B_V ～0.01 T) was found to narrow the plasma density and boron deposition profiles near the UH resonance, thus better localizing the deposition. A Monte Carlo simulation is developed to model the boron deposition on the different QMB/tokamak surfaces. The model requires a relatively high boron ion gyroradius of ～5 mm, indicating a B⁺¹ ion temperature of ～2 eV, to match the deposition on QMB surfaces with different orientation to B_T. Additionally, the boron ion trajectories become de-magnetized at high neutral gas throughput (～0.5 Pa m³ s^?1) and pressure (～2 Pa) when the largest absolute deposition rates are measured, resulting in deposition patterns, which are independent of surface orientation to B_T in optimized conditions.     

The ITER full size plasma source device design

In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full size plasma source with low voltage extraction and a full size NB injector at full beam power (1 MV). These two different devices will separately address the main scientific and technological issues of the 17 MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1 h. The required negative ion current density to be extracted from the plasma source ranges from 290 A/m² in D₂ (D^?) and 350 A/m² in H₂ (H^?) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3 Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1 m².The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.     

Characterization of plasma jet ejected from a parallel-plate rail gun for simulating edge localized mode

A small plasma gun with parallel-plate configuration is fabricated to generate a bunch of plasma which is similar to ELM (edge localized mode) plasma, by taking advantages of its simplicity and cost-effectiveness. Prior to explore how to control the ELM-like plasma so as to relieve heat load on the divertor target, characteristics of a plasma jet ejected from the plasma gun are investigated using a quadruple Langmuir probe which is appropriate for measuring rapidly varying plasma parameters such as electron density, temperature, and ion velocity at the same time. The plasma density and ion velocity measured at 112 mm away from the exit are 3 × 10¹⁹ m^?3 and 11 km/s, respectively, which seem to be suitable for investigating next step research on the control of ELM-like plasma using various methods such as electromagnetic waves and high-voltage pulses. Also, the quadruple Langmuir probe is proven to be adequate for use in such experiments.     

Gaseous radiocarbon measurements of small samples

Radiocarbon dating by means of accelerator mass spectrometry (AMS) is a well-established method for samples containing carbon in the milligram range. However, the measurement of small samples containing less than 50 μg carbon often fails. It is difficult to graphitise these samples and the preparation is prone to contamination. To avoid graphitisation, a solution can be the direct measurement of carbon dioxide. The MICADAS, the smallest accelerator for radiocarbon dating in Zurich, is equipped with a hybrid Cs sputter ion source. It allows the measurement of both, graphite targets and gaseous CO₂ samples, without any rebuilding.This work presents experiences dealing with small samples containing 1-40 μg carbon. 500 unknown samples of different environmental research fields have been measured yet. Most of the samples were measured with the gas ion source. These data are compared with earlier measurements of small graphite samples. The performance of the two different techniques is discussed and main contributions to the blank determined. An analysis of blank and standard data measured within years allowed a quantification of the contamination, which was found to be of the order of 55 ng and 750 ng carbon (50 pMC) for the gaseous and the graphite samples, respectively. For quality control, a number of certified standards were measured using the gas ion source to demonstrate reliability of the data.     

Optimised plasma stabilisation at TEXTOR with an advanced, real-time digital control scheme

The tokamak TEXTOR at the Research Centre in Jülich is in operation since more than 25 years. The various control systems at the start, in 1982, were based on analogue techniques, a standard at the time, and were later partly replaced by specially developed digital systems. These systems proved their robustness over the years. As a replacement for the old system, off-the-shelf products were used to ensure continuity, reliability and to reduce the development cost. To provide advanced control scenarios, the new system allows the implementation of more sophisticated algorithms for magnetic and kinetic control. The LabVIEW Real-Time (RT) modules and real-time hardware from National Instruments satisfy these requirements to a large extent. The new system has already been successfully commissioned at TEXTOR and is used to calculate in real-time the plasma density profile (10 ms), the Shafranov shift (10 ms), the plasma vertical and horizontal position (20 $\mu$s) and to control the plasma shape (1 ms). TEXTOR has circular plasmas and has an iron core. Its central part is operated in saturation. During the saturation phase, stray fields change the plasma shape from nearly circular to slightly triangular. By using a shape-control coil set, we can control and adjust the plasma form. The new real-time system is presented as well as the implemented control applications.     

Negative ion test facility ELISE—Status and first results

The new test facility ELISE (Extraction from a Large Ion Source Experiment) has been designed and installed since November 2009 at IPP Garching to support the development of the radio frequency driven negative ion source for the Neutral Beam System on ITER. The test facility is now completely assembled; all auxiliary systems have been commissioned and are operational. First plasma and beam operation is starting in October 2012.The source is designed to deliver an ion beam of 20 A of D^? ions, operating at 0.3 Pa source pressure at an electron to ion current ratio below 1. Beam extraction is limited to 60 kV for 10 s every 3 minutes, while plasma operation of the source can be performed continuously for 1 hour. The ion source and extraction system have the same width as the ITER source, but only half the height, i.e. 1 × 1 m² source area with an extraction area of 0.1 m². The aperture pattern of the extraction system and the multi driver source concept stay as close as possible to the ITER design. Easy access to the source for diagnostic tools or modifications allows to analyze and optimize the source performance. Among other possibilities many different magnetic filter field configurations inside the source can be realized to enhance the negative ion extraction and to reduce the co-extraction of electrons. Beam power and profiles are measured by calorimetry and thermography on an inertially cooled target as well as by beam emission spectroscopy. Cs evaporation into the source is done via two dispenser ovens.     

Development of a plasma driven permeation experiment for TPE

Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1], [2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes.     

Recent and future upgrades to the DIII-D tokamak

Research on the DIII-D tokamak focuses on support for next-generation devices such as ITER by providing physics solutions to key issues and advancing the fundamental understanding of fusion plasmas. To support this goal, the DIII-D facility is planning a number of upgrades that will allow improved plasma heating, control, and diagnostic measurement capabilities. The neutral beam system has recently added an eighth ion source and one of the beamlines is currently being rebuilt to allow injection of 5 MW of off-axis power at an angle of up to 16.5° from the horizontal. The electron cyclotron heating (ECH) system is adding two additional gyrotrons and is using new launchers that can be aimed poloidally in real-time by an improved plasma control system. The fast wave heating system is being upgraded to allow two of the three launchers to inject up to 2 MW each in future experiments. Several diagnostics are being added or upgraded to more thoroughly study fluctuations, fast ions, heat flux to the walls, plasma flows, rotation, and details of the plasma density and temperature profiles.     

Near-infrared detectors for ITER LIDAR Thomson scattering

In this paper we discuss strategies for the development of fast photodetectors suitable for operation in the λ > 850 nm near-infrared (NIR) spectral region in the ITER core LIDAR Thomson scattering (TS) system. Detection of this spectral range is necessary if a Nd:YAG laser operating at the fundamental wavelength (λ = 1.06 μm) will be used as the input laser source. Different types of NIR photodetectors are potentially suitable for use in ITER LIDAR TS: the transferred electron (TE) InGaAsP/InP hybrid photodiodes and microchannel plate photomultipliers (MCP PMTs), the In_xGa_1?xAs MCP image intensifiers and PMTs, and the detectors based on transmission Si photocathodes. But their characteristics of either sensitivity, active area or speed of response, do not match the ITER specifications and all devices require some developmental work. For each of these detector types we review the characteristics of devices presently available and suggest a realistic development strategy suitable to extend their performances to meet the ITER specifications. Finally the expected performance of the ITER LIDAR TS system for different detector choices are compared by calculating the expected signal-to-noise ratio of the measured plasma temperature and density.     

Development of a high energy pulsed plasma simulator for the study of liquid lithium trenches

To simulate detrimental events in a tokamak and provide a test-stand for a liquid-lithium infused trench (LiMIT) device [1], a pulsed plasma source utilizing a theta pinch in conjunction with a coaxial plasma accelerator has been developed. The plasma is characterized using a triple Langmuir probe, optical methods, and a calorimeter. Clear advantages have been observed with the application of a coaxial plasma accelerator as a pre-ionization source. The experimental results of the plasma gun in conjunction with the existing theta pinch show a significant improvement from the previous energy deposition by a factor of 14 or higher, resulting in a maximum energy and heat flux of 0.065 ± 0.002 MJ/m² and 0.43 ± 0.01 GW/m². A few ways to further increase the plasma heat flux for LiMIT experiments are discussed.     

Fueling requirements of super-high-density plasmas towards innovative ignition regime

Super-high-density plasma with an internal diffusion barrier which is observed in the Large Helical Device has been extrapolated to a fusion reactor grade plasma to explore an innovative ignition regime and to clarify essential requirements for pellet fueling. The peaked density profiles due to the internal diffusion barrier formation allow reduction in the required minimum temperature to sustain a self-burning plasma down to 10 keV. Direct core fueling beyond the internal diffusion barrier is essential to sustain the peaked density profile. Furthermore, the acceptable pellet size becomes small in terms of fusion output perturbation because the effective volume of the burning plasma becomes small with the peaked density profile. In order to sustain a self-burning plasma with an internal diffusion barrier, therefore, extremely high velocity pellet injection beyond 10 km/s is inevitable unless another solution to the core fueling is found.     

Influence of temperature and plasma composition on deuterium retention in refractory metals

Refractory materials are being considered potential candidates to build the first wall of the fusion reactor chamber. This work reports on the results of the study of tungsten and molybdenum metals exposed to high flux densities (～10²⁴ D/m² s) and low temperature (T_e ～ 3 eV) deuterium plasmas in Pilot-PSI irradiation facility.The hydrogenic retention in poly-crystalline W and Mo targets was studied with ³He nuclear reaction analyses (NRA). The NRA results clearly show a two-dimensional radial distribution of the deuterium with a minimum at the center and a maximum close to the edge. These distribution correlates well with the thermal profile of the sample surface, where a maximum of ～1600 K was measured at the center decreasing to ～1000 K in the edges. A maximum deuterium fluence retention of 5 × 10¹⁵ D/cm² was measured. The values of the retained fractions ranging from 10^?5 to 10^?6 D_retained/D_incident were measured with thermal desorption spectroscopy (TDS) and compares well with IBA results. Moreover, the presence of C in the plasma and its co-deposition increases the D retention in the region where a C film is formed. Both NRA and TDS results show no clear dependence of retention on incident fluence suggesting the absence of plasma related traps in W under these conditions.     

Hardening of self ion implanted tungsten and tungsten 5-wt% rhenium

Tungsten is one of the most promising materials for high temperature components in any future nuclear fusion tokamak. In this study tungsten-ion implantation has been used to simulate the damage caused by neutrons in pure tungsten and tungsten 5 wt% rhenium. This damaged layer is only 300 nm deep so conventional mechanical tests cannot be used to investigate it. Nanoindentation has been used to measure the change in hardness as a function of six damage levels (0 dpa, 0.07 dpa, 0.4 dpa, 1.2 dpa, 13 dpa and 33 dpa). In pure tungsten the hardness increase is seen to saturate by 0.4 dpa at ≈0.8 GPa. Transmission electron microscopy of the damage structure sees a similar saturation of the loop volume number density at the same damage level. In the tungsten 5 wt% rhenium the increase in hardness is constant between 0.07 and 1.2 dpa, ≈0.85 GPa. The loop volume number density as measured using TEM is also shows little change in this region. At a damage level of 33 dpa the hardness increase is 2.88 GPa; this corresponds with the formation of small 3–5 nm rhenium clusters as observed using atom probe tomography.     

Operational status of the Magnum-PSI linear plasma device

The construction phase of the linear plasma generator Magnum-PSI at the FOM institute DIFFER has been completed and the facility has been officially opened in March 2012. The scientific program to gain more insight in the plasma–wall interactions relevant for ITER and future fusion reactors has started.In Magnum-PSI, targets of a wide range of materials and shapes can be exposed to high particle, high heat flux plasmas (>10²⁴ ions m^?2 s^?1; >10 MW/m²). For magnetization of the plasma, oil-cooled electromagnets are temporarily installed to enable pulsed operation until the device is upgraded with a superconducting magnet. The magnets generate a field of up to 1.9 T close to the plasma source for a duration of 6 s. Longer exposure times are available for lower field settings.Plasma characterizations were done with a variety of gases (H, D, He, Ne and Ar) to determine the machine performance and prepare for subsequent scientific experiments. Thomson scattering and optical emission spectroscopy were used to determine the plasma parameters while infrared thermography and target calorimetry were used to determine the power loads to the surface.This paper reports on the status of Magnum-PSI and its diagnostic systems. In addition, an overview of the plasma parameters that can be achieved in the present state will be given.     

Development of a TOF-ERDA measurement system for analysis of light elements using a He beam

A time-of-flight ERDA (TOF-ERDA) measurement system has been developed for the analysis of light elements. He ions are used for the incident beam, and recoil light ions are detected with the system. The system consists of a time detector and a silicon detector, and energy and velocity of recoil ion are measured simultaneously. The depth resolution of 21.6 ± 2.2 nm (FWHM) has been obtained by an ERDA measurement of a thin carbon layer onto a silicon wafer using a 5.7 MeV He beam. The mass resolution is better than 1 for elements up to oxygen. Maximum detectable depth of carbon in a PET film is about 650 nm. An ERDA measurement of implanted carbon in a silicon wafer has been demonstrated.     

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.     

Operational characteristics of the high flux plasma generator Magnum-PSI

In Magnum-PSI (MAgnetized plasma Generator and NUMerical modeling for Plasma Surface Interactions), the high density, low temperature plasma of a wall stabilized dc cascaded arc is confined to a magnetized plasma beam by a quasi-steady state axial magnetic field up to 1.3 T. It aims at conditions that enable fundamental studies of plasma–surface interactions in the regime relevant for fusion reactors such as ITER: 10²³–10²⁵ m⁻² s⁻¹ hydrogen plasma flux densities at 1–5 eV. To study the effects of transient heat loads on a plasma-facing surface, a high power pulsed magnetized arc discharge has been developed. Additionally, the target surface can be transiently heated with a pulsed laser system during plasma exposure. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.     

First-principles study of hydrogen adsorption and permeation in reconstructed cubic erbium oxide surfaces

Understanding surface properties of Er₂O₃, especially in relation to adsorption and permeation of atomic hydrogen, is of considerable importance to the study of tritium permeation barriers. In this work, hydrogen diffusion pathways through the low-index (1 0 0), (1 1 0), and (1 1 1) surfaces of cubic Er₂O₃ have been calculated using density functional theory within the GGA (PBE) + U approach. The dependence of the effective U parameter on lattice constants, bulk moduli, and formation energies of Er₂O₃ has been investigated in detail. The energetics of hydrogen penetration from the surfaces to the solution site in bulk Er₂O₃ were defined using the optimum effective U value of 5.5 eV. For a low surface coverage of hydrogen (0.89 × 10¹⁴ H/cm²), a penetration energy of at least 1.7 eV was found for all the low-index erbium oxide surfaces considered. The results of the present study will provide useful guidance for future studies on modeling defects, such as grain boundaries and vacancies, in tritium permeation barriers.