The Prototype Material Plasma Exposure eXperiment (Proto-MPEX) is a pulsed linear plasma device that uses a radio frequency (RF) helicon antenna and a helicon ceramic window as the plasma ionization source. Modeling performed within this work on the helicon operations predicts hot spots on the helicon ceramic window due to an RF-induced sheath. The RF sheath potential leads to physical sputtering of the ceramic window predominantly where the helicon antenna sits around the helicon window. The role of oxygen in sputtering within Proto-MPEX is also investigated and is needed to match the experimental data. Experiments on Proto-MPEX show reasonable agreement with the modeled erosion/redeposition pattern on the helicon window, and the modeled impurity flux to the target shows radial transport within the device as expected from classical collisional impurity transport. Temperature screening of impurities is shown to be needed for Proto-MPEX to reduce impurity flux to the target. The modeling approach used in this work can be applied to other plasma RF applications for the determination of impurity production and transport.

•New linear plasma device MPEX will address plasma material interaction gaps for fusion.•MPEX will have increased capabilities over existing linear plasma devices and will be world-leading when completed.•MPEX will expose apriori neutron irradiated material samples for fusion reactor divertor relevant heat and particle fluxes.•The conceptual design of MPEX is finished.

Industrial applications of additively manufactured components are increasing quickly. Adequate quality control of the parts is necessary in ensuring safety when using these materials. Base material properties, surface conditions, as well as location and size of defects are some of the main targets for nondestructive evaluation of additively manufactured parts, and the problem of adequate characterization is compounded given the challenges of complex part geometry. Numerical modeling can allow the interplay of the various factors to be studied, which can lead to improved measurement design. This paper presents a finite element simulation verified by experimental results of ultrasonic waves scattering from flat bottom holes (FBH) in additive manufacturing materials. A focused beam immersion ultrasound transducer was used for both the modeling and simulations in the additive manufactured samples. The samples were SS17 4 PH steel samples made by laser sintering in a powder bed.

The Material Plasma Exposure eXperiment (MPEX) is a linear plasma device planned to address plasma-material interactions for future fusion reactors. Its concept foresees the capability to expose apriori neutron irradiated material samples to fusion reactor grade divertor plasmas. This new capability will be unique worldwide addressing important research needs in the area of fusion nuclear science. It will be an evolution to current operating steady-state linear plasma devices, which are limited either in plasma fluxes they can deliver to the material targets or plasma temperatures (for ions and electrons) they can reach in front of the material targets. The concept of MPEX foresees a combination of a high-power helicon plasma source with microwave electron heating and ion cyclotron resonance heating. This source and heating concept is being tested on the Prototype-Material Plasma Exposure eXperiment (Proto-MPEX). With 100 kW helicon power, a plasma density of 8 x 1019 m-3 was achieved, which is about a factor two more than required for MPEX. Electron heating was pursued with a 28 GHz gyrotron. A maximum power of 50 kW was delivered to the plasma, which is produced by the helicon. At this frequency, the plasma is overdense in the plasma center (> 1 x 1019 m-3). Maximum electron temperatures of 20 eV have been achieved under those overdense plasma conditions with Electron Bernstein Wave (EBW) heating. This is almost the electron temperature required for MPEX (25-30 eV). Ion cyclotron heating (ICH) was performed in the frequency range of 6 – 12 MHz with a low power ICH antenna able to launch about 25-30 kW of power. Without ICH, the ion temperature is about 2-4 eV. With ICH, ion temperatures of 8-12 eV were measured. The ion fluxes to the target are about 5 x 1023 m-2s-1. The plasmas produced by the helicon antenna have been modeled extensively with a fluid plasma code, coupled to a Monte-Carlo neutral code (B2-Eirene). The plasma transport can be well explained by this fluid approach and a radial diffusion coefficient consistent with Bohm-like transport.

The Prototype-Material Plasma Exposure eXperiment (Proto-MPEX) is a linear plasma device to test the plasma source concepts for the future Material Plasma Exposure eXperiment (MPEX). MPEX is planned to address plasma material interactions under fusion reactor divertor conditions as will occur in devices like DEMO, including the plasma exposure of a-priori neutron irradiated materials. This requires a novel linear plasma device with a plasma source that can produce high electron densities along with heating systems to obtain high electron and ion temperatures. In Proto-MPEX a high-power helicon source was tested together with electron heating (28 GHz microwaves) and ion cyclotron heating (8.5-12 MHz). Electron densities in excess of 1 x 10(20) m(-3) were produced with the helicon antenna. Electron temperatures can reach 20 eV and are typically below 6 eV at high densities (> 5 x 10(19) m(-3)). Ion heating has been demonstrated at low electron densities (<2 x 10(19) m(-3)) with Ion temperatures of up to 13 eV. Maximum heat fluxes of 14 MW/m(2) have been obtained in low density discharges (similar to 5 x 10(18) m(-3)), while heat fluxes are typically only 2-3 MW/m(2) in high density discharges (> 2 x 10(19) m(-3)).

The availability of future fusion devices such as a Fusion Nuclear Science Facility (FNSF) or DEMO greatly depends on long operating lifetimes of plasma facing components in their divertors. ORNL is designing the Material-Plasma Exposure eXperiment (MPEX), a superconducting magnet, steady-state device to address the plasma material interactions of fusion reactors. MPEX will utilize a new high-intensity plasma source concept based on RF technology. This source concept will allow the experiment to cover the entire expected plasma conditions in the divertor of a future fusion reactor. It will be able to study erosion and re-deposition for relevant geometries with relevant electric and magnetic fields in-front of the target. MPEX is being designed to allow for the exposure of a-priori neutron-irradiated samples. The target transfer cask has been designed to undock from the linear plasma generator such that it can be transferred to diagnostics stations for more detailed surface analysis. MPEX is being developed in a staged approach with successively increased capabilities. After the initial development step of the helicon source and ECH system the source concept is being tested in the Proto-MPEX device (100 kW helicon, 200 kW EBW, 30 kW ICRH). Proto-MPEX has achieved electron densities of more than 4×1019m−3 with a large diameter (13cm) helicon antenna at 100 kW power. First heating with microwaves resulted in a higher ionization represented by higher electron densities on axis, when compared to the helicon plasma only without microwave heating.

The non-solenoid start-up technique reported here relies on a double mode conversion for electron Bernstein wave (EBW) excitation. It consists of the mode conversion of the ordinary mode, entering the plasma from the low field side of the tokamak, into the extraordinary (X) mode at a mirror-polarizer located at the high field side. The X mode propagates back to the plasma, passes through electron cyclotron resonance and experiences a subsequent X to EBW mode conversion near the upper hybrid resonance. Finally the excited EBW mode is totally absorbed at the Doppler shifted electron cyclotron resonance. The absorption of EBW remains high even in cold rarefied plasmas. Furthermore, EBW can generate significant plasma current giving the prospect of a fully solenoid-free plasma start-up. First experiments using this scheme were carried out on MAST [V. Shevchenko et al, Nuclear Fusion 50, 022004 (2010)]. Plasma currents up to 33 kA have been achieved using 28 GHz 100kW 90ms RF pulses. Recently experimental results were extended to longer RF pulses showing further increase of plasma currents generated by RF power alone. A record current of 73kA has been achieved with 450ms RF pulse of similar power. The current drive enhancement was mainly achieved due to RF pulse extension and further optimisation of the start-up scenario.

Research on the National Spherical Torus Experiment, NSTX, targets physics understanding needed for extrapolation to a steady-state ST Fusion Nuclear Science Facility, pilot plant, or DEMO. The unique ST operational space is leveraged to test physics theories for next-step tokamak operation, including ITER. Present research also examines implications for the coming device upgrade, NSTX-U. A E scaling appropriate for varied wall conditions exhibits a strong improvement of BTE with decreased electron collisionality produced by lithium (Li) wall conditioning. Nonlinear microtearing simulations match experimental electron diffusivity quantitatively and predict reduced electron heat transport at lower collisionality. Beam-emission spectroscopy measurements indicate the poloidal correlation length of pedestal turbulence ~ 10 i increases at higher electron density gradient and lower Ti gradient. Plasma characteristics change nearly continuously with increasing Li evaporation and ELMs stabilize due to edge density gradient alteration. Global mode stability studies show stabilizing resonant kinetic effects are enhanced at lower collisionality. Combined radial and poloidal field sensor feedback controlled n = 1 perturbations and improved stability. The disruption probability due to unstable RWMs is reduced at high N/li > 11 consistent with low frequency MHD spectroscopy measurements of mode stability. Greater instability seen at intermediate N is consistent with decreased kinetic RWM stabilization. A model-based RWM state-space controller produced long-pulse discharges exceeding N = 6.4 and N/li = 13. Precursor analysis shows 98% of disruptions can be predicted with 10ms warning and a false positive rate of only 6%. Disruption halo currents rotate toroidally and can have significant toroidal asymmetry. Global kinks cause measured fast ion redistribution. Full-orbit calculations show redistribution from the core outward and toward V||/V = 1 where destabilizing CAE resonances are expected. Applied 3D fields alter GAE characteristics. The snowflake divertor configuration enhanced by radiative detachment shows large reductions in both steadystate and ELM heat fluxes (steady-state peak values down from 7 MW/m to less than 1 MW/m). Toroidal asymmetry of heat deposition is observed during ELMs or by 3D fields. Coaxial helicity injection has reduced the inductive startup flux, with plasmas ramped to 1MA requiring 35% less inductive flux. Non-inductive current fraction (NICF) up to 65% is reached experimentally with NBI at Ip = 0.7 MA and between 70 – 100% with high harmonic fast wave application at Ip = 0.3 MA. NSTX-U scenario development calculations project 100% NICF for a large range of 0.6 < Ip (MA) < 1.35.

An engineering study has been poformed on the ITER electron cyclotron transmission lines with the aim of optimizing its conceptual design. The support types and optimum spacing, cooling, vacuum, seismic, and gravitational effects were reviewed. For the vacuum system it was shown that two pumps per line, with a capacity of 50 l/s, are sufficient. It was explained that the temperature variation inside the building is the predominant factor that influences the thermal expansion of the lines. The support strategy is one of minimizing the number of constraints. Variation in support interspacing reduces the degree of harmonic disturbances. The section of transmission line inside the ITER port cell was identified as critical with regards to occurrence of deformation and stresses. Potential solutions are described. The use of seismic breaks is discussed in light of the differences in foundation and structure of the ITER tokamak building and assembly hall. It is proposed that this interface be studied in more detail, after more data is available on the behavior of these buildings. The geometry of individual supports should be simple, with the fewest possible adjustments. The supports are designed to allow small movements of the waveguide to compensate for the thermal expansion or contraction. The transmission line system can be made for optimum alignment during nominal operating temperatures by prestressing during installation.

The ITER ECH&CD system will have an installed power of 24 MW at 170 GHz delivered to the launchers at the plasma side through a transmission line sub-system constituted by evacuated HE11 waveguides, DC breaks, power monitors, mitre bends, polarizers, switches, loads and pumping sections. Each line will be typically about 160 m in length and will connect the RF power sources alternatively to the equatorial launcher or to one of the upper launchers in order to accomplish the various physics requirements: heating, current drive and instability control. Two different source configurations are at the moment under study, pending the fmal decision about the European 2 MW coaxial gyrotron development. Each configuration will imply a dedicated transmission line layout. An overview of the actual design is presented and the technical requirements are discussed.

A 24MW CW Electron Cyclotron Heating and Current Drive (EC H&CD) system operating at 170GHz is to be installed for the ITER tokamak. The EC system will represent a large step forward in the use of microwave systems for plasma heating for fusion applications; present day systems are operating in relatively short pulses (<= 10s) and installed power levels of <= 4.5MW. The magnitude of the ITER system necessitates a worldwide collaboration. This is also reflected in the EC system that is comprised of the power supplies, sources, transmission line and launchers. A partnership between Europe, India, Japan, Russia, United States and the ITER organization is formed to collaborate on design and R&D activities leading to the procurement, installation, commissioning and operation of this system.

A 24MW CW Electron Cyclotron Heating and Current Drive (EC H&CD) system operating at 170GHz is to be installed for the ITER tokamak. The EC system will represent a large step forward in the use of microwave systems for plasma heating for fusion applications; present day systems are operating in relatively short pulses (≤10s) and installed power levels of ≤4.5MW. The magnitude of the ITER system necessitates a worldwide collaboration. This is also reflected in the EC system that is comprised of the power supplies, sources, transmission line and launchers. A partnership between Europe, India, Japan, Russia, United States and the ITER organization is formed to collaborate on design and R&D activities leading to the procurement, installation, commissioning and operation of this system. The aim of this paper is to provide a brief review of the design improvements incorporated in 2008 and 2009 following the ITER design review of 2007. The improvements aim at increased functional capabilities, simplification of interfaces and reduction of procurement and operating risks.

A 26MW Electron Cyclotron Heating and Current Drive (EC H&CD) system is to be installed for ITER. The main objectives are to provide, start-up assist, central H&CD and control of MHD activity. These are achieved by a combination of two types of launchers, one located in an equatorial port and the second type in four upper ports. The physics applications are partitioned between the two launchers, based on the deposition location and driven current profiles. The equatorial launcher (EL) will access from the plasma axis to mid radius with a relatively broad profile useful for central heating and current drive applications, while the upper launchers (ULs) will access roughly the outer half of the plasma radius with a very narrow peaked profile for the control of the Neoclassical Tearing Modes (NTM) and sawtooth oscillations. The EC power can be switched between launchers on a time scale as needed by the immediate physics requirements. A revision of all injection angles of all launchers is under consideration for increased EC physics capabilities while relaxing the engineering constraints of both the EL and ULs. A series of design reviews are being planned with the five parties (EU, IN, JA, RF, US) procuring the EC system, the EC community and ITER Organization (10). The review meetings qualify the design and provide an environment for enhancing performances while reducing costs, simplifying interfaces, predicting technology upgrades and commercial availability. In parallel, the test programs for critical components are being supported by 10 and performed by the Domestic Agencies (DAs) for minimizing risks. The wide participation of the DAs provides a broad representation from the EC community, with the aim of collecting all expertise in guiding the EC system optimization. Still a strong relationship between 10 and the DA is essential for optimizing the design of the EC system and for the installation and commissioning of all ex-vessel components when several teams from several DAs will be involved together in the tests on the ITER site.