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Harrison, J.

Publications and source records attributed to Harrison, J..

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Real-time plasma equilibrium reconstruction and shape control for the MAST Upgrade tokamak

Real-time magnetic control has been developed to deliver precise control of multiple plasma shape parameters for advanced divertor configurations, including double-null, Super-X, X-point target and X-divertor for the first time on the MAST Upgrade (MAST-U) spherical tokamak. Successful real-time magnetic equilibrium control of different plasma shape variables has been accomplished in the 2022–2023 MAST-U experimental campaign for the advanced MAST-U divertor configurations. Application of the MAST-U boundary reconstruction algorithm, LEMUR, is described and compared with off-line equilibrium reconstruction and diagnostic measurements. The process of design and verification of the axisymmetric magnetic control schemes using a suite of control analysis tools (known collectively as TokSys) is also described.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

The Digital Mirror Langmuir Probe: Field Programmable Gate Array Implementation of Real-Time Langmuir Probe Biasing

High bandwidth, high spatial resolution measurements of electron temperature, density and plasma potential are valuable for resolving turbulence in the boundary plasma of tokamaks. While conventional Langmuir probes can provide such measurements either their temporal or spatial resolution is limited: the former by the sweep rate necessary for obtaining I-V characteristics and the latter by the need to use multiple electrodes, as is the case in triple and double probe configurations. The Mirror Langmuir Probe (MLP) bias technique overcomes these limitations by rapidly switching the voltage on a single electrode cycling between three bias states, each dynamically optimised for the local plasma conditions. The MLP system on Alcator C-Mod used analog circuitry to perform this function, measuring Te, VF, and Isat at 1.1 MSPS. Recently, a new prototype digital MLP controller has been implemented on a Red Pitaya Field Programmable Gate Array (FPGA) board which reproduces the functionality of the original controller and performs all data acquisition. There is also the potential to provide the plasma parameters externally for use with feedback control systems. The use of FPGA technology means the system is readily customisable at a fraction of the development time and implementation cost. A second Red Pitaya was used to test the MLP by simulating the current response of a physical probe using C-Mod experimental measurements. This project is available as a git repository to facilitate extensibility (e.g. real-time control outputs, more voltage states) and scalability through collaboration.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Overview of the TCV tokamak program: Scientific progress and facility upgrades

The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Demonstration of improvement in 2D tomographic reconstructions of filtered cameras in the JET ITER-like wall divertor due to corrected interpretation of reflections

Improved localization of emission and reduction of artefacts have been achieved in tomographic reconstructions of the tangentially viewing spectroscopic divertor cameras in JET by introducing the contribution of reflections from the metallic wall surfaces in the geometry definition of the tomography process. The reconstructions thus describe the plasma emission only without features arising from misinterpreted reflected light. Benchmarks with synthetic emission distributions showed accurate reproduction within 10% and 1–2% of the reference emission distributions and camera images, respectively, when the reflection correction was applied. Without the correction, the reconstructed emission was observed to concentrate from the divertor volume towards the wall surfaces together with bright emission artefacts appearing in front of the surfaces to reproduce the reflections seen in the camera images.

Karhunen, Juuso↗
DOE OSTI2019

The digital mirror Langmuir probe: Field programmable gate array implementation of real-time Langmuir probe biasing

High bandwidth, high spatial resolution measurements of electron temperature, density, and plasma potential are valuable for resolving turbulence in the boundary plasma of tokamaks. While conventional Langmuir probes can provide such measurements, either their temporal or spatial resolution is limited: the former by the sweep rate necessary for obtaining I-V characteristics and the latter by the need to use multiple electrodes, as is the case in triple and double probe configurations. The Mirror Langmuir Probe (MLP) bias technique overcomes these limitations by rapidly switching the voltage on a single electrode cycling between three bias states, each dynamically optimized for the local plasma conditions. The MLP system on Alcator C-Mod used analog circuitry to perform this function, measuring Te, VF, and Isat at 1.1 MSPS. Recently, a new prototype digital MLP controller has been implemented on a Red Pitaya Field Programmable Gate Array (FPGA) board which reproduces the functionality of the original controller and performs all data acquisition. There is also the potential to provide the plasma parameters externally for use with feedback control systems. The use of FPGA technology means the system is readily customizable at a fraction of the development time and implementation cost. A second Red Pitaya was used to test the MLP by simulating the current response of a physical probe using C-Mod experimental measurements. This project is available as a git repository to facilitate extensibility (e.g., real-time control outputs and more voltage states) and scalability through collaboration.

47 OTHER INSTRUMENTATION↗
DOE OSTI2018

On the role of finite grid extent in SOLPS-ITER edge plasma simulations for JET H-mode discharges with metallic wall

The impact of the finite grid size in SOLPS-ITER edge plasma simulations is assessed for JET H-mode discharges with a metal wall. For a semi-horizontal divertor configuration it is shown that the separatrix density is at least 30% higher when a narrow scrape-off layer (SOL) grid width is chosen in SOLPS-ITER compared to the case for which the SOL grid width is maximised. The density increase is caused by kinetic neutrals being not confined inside the divertor region because of the reduced extent of the plasma grid. In this case, an enhanced level of reflections of energetic neutrals at the low-field side (LFS) metal divertor wall is observed. This leads to a shift of the ionisation source further upstream which must be accounted for as a numerical artefact. An overestimate in the cooling at the divertor entrance is observed in this case, identified by a reduced heat flux decay parameters λ q div . Otherwise and further upstream the mid-plane heat decay length λ q parameter is not affected by any change in divertor dissipation. This confirms the assumptions made for the ITER divertor design studies, i.e. that λ q upstream is essentially set by the assumptions for the ratio radial to parallel heat conductivity. It is also shown that even for attached conditions the decay length relations λ ne > λ Te > λ q hold in the near-SOL upstream. Thus for interpretative edge plasma simulations one must take the (experimental) value of λ ne into account, rather than λ q , as the former actually defines the required minimum upstream SOL grid extent.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Investigation into the formation of the scrape-off layer density shoulder in JET ITER-like wall L-mode and H-mode plasmas

The low temperature boundary layer plasma (scrape-off layer or SOL) between the hot core and the surrounding vessel determines the level of power loading, erosion and implantation of material surfaces, and thus the viability of tokamak-based fusion as an energy source. Here, this study explores mechanisms affecting the formation of flattened density profiles, so-called 'density shoulders', in the low-field side (LFS) SOL, which modify ion and neutral fluxes to surfaces—and subsequent erosion. There is evidence against local enhancement of ionization inducing shoulder formation. We find that increases in SOL parallel resistivity, Λ div (=[ L || ν ei Ω i ]/ c s Ω e ), postulated to lead to shoulder growth through changes in SOL turbulence characteristics, correlates with increases in SOL shoulder amplitude, A s , but only under a subset of conditions (D 2 -fuelled L-mode density scans with outer strike point on the horizontal target). Λ div fails to correlate with A s for cases of N 2 seeding or during sweeping of the strike point across the horizontal target. The limited correlation of Λ div and A s is also found for H-mode discharges. Thus, while it may be necessary for Λ div to be above a threshold of ~1 for shoulder formation and/or growth, another mechanism is required. More significantly, we find that in contrast to parallel resistivity, outer divertor recycling, as quantified by the total outer divertor Balmer D α emission, I – D α , does scale with A s where Λ div does and even where Λ div does not. Divertor recycling could lead to SOL density shoulder formation through: (a) reducing the parallel to the field flow (loss) of ions out of the SOL to the divertor; and (b) changes in radial electric fields which lead to E × B poloidal flows as well as potentially affecting SOL turbulence birth characteristics. Therefore, changes in divertor recycling may be the sole process involved in bringing about SOL density shoulders or it may be that it acts in tandem with parallel resistivity.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Expanding the role of impurity spectroscopy for investigating the physics of high-Z dissipative divertors

New techniques that attempt to more fully exploit spectroscopic diagnostics in the divertor and pedestal region during highly dissipative scenarios are demonstrated here using experimental results from recent low-Z seeding experiments on Alcator C-Mod, JET and ASDEX Upgrade. To exhaust power at high parallel heat flux, $q_∥$ > 1 GW/m 2 , while minimizing erosion, reactors with solid, high-Z plasma facing components (PFCs) are expected to use extrinsic impurity seeding. Due to transport and atomic physics processes which impact impurity ionization balance, so-called ‘non-coronal’ effects, we do not accurately know and have yet to demonstrate the maximum $q_∥$ which can be mitigated in a tokamak. Radiation enhancement for nitrogen is shown to arise primarily from changes in Li- and Be-like charge states on open field lines, but also through transport-driven enhancement of H- and He-like charge states in the pedestal region. Measurements are presented from nitrogen seeded H-mode and L-mode plasmas where emission from N 1+ through N 6+ are observed. Active charge exchange spectroscopy of partially ionized low-Z impuri- ties in the plasma edge is explored to measure N 5+ and N 6+ within the confined plasma, while passive UV and visible spectroscopy is used to measure N 1+ -N 4+ in the boundary. Examples from recent JET and Alcator C-Mod experiments which employ nitrogen seeding highlight how improving spectroscopic coverage can be used to gain empirical insight and provide more data to validate boundary simulations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗