The New PEGASUS-III Experiment

Sontag, A. C.; Fonck, R. J.; Borchardt, Michael T.; Diem, S. J.; Fonck, R. J.; Keyhani, Armand K.; Kujak-Ford, B.A.; Lewicki, B.T.; Nornberg, M. D.; Palmer, Alan C.; Pierren, C.; Reusch, J.A.; Rhodes, A.T.; Richner, N.J.; Rodriguez-Sanchez, Cuauhtemoc; Schaefer, C.; Weberski, Justin D.; Winz, G.R.

doi:10.1109/tps.2022.3184626

Cited by 4 publications

(2 citation statements)

References 11 publications

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“…To address the startup issue, the PEGASUS-III Experiment (see Fig. 1) will provide a dedicated U.S. platform for solenoid-free startup development [2], [3]. This is a major upgrade to the PEGASUS facility located at the University of Wisconsin at Madison, Madison, WI, USA, that is presently underway.…”

Section: Introductionmentioning

confidence: 99%

A Coaxial Helicity Injection System for Nonsolenoidal Startup Studies on the PEGASUS-III Experiment

Reusch

Raman

Bongard

et al. 2022

IEEE Trans. Plasma Sci.

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Initiating current without using magnetic induction from a central solenoid is a critical scientific and technical challenge facing the spherical tokamak (ST). One such technique that has shown promise on several devices is coaxial helicity injection (CHI). In CHI, a dc voltage applied to large area coaxial electrodes injects current and helicity into the vacuum vessel for plasma startup and plasma current (I p ) sustainment. Major outstanding issues for CHI include eliminating the need for a vacuum vessel break; the scaling of I p with injector and/or flux footprint shape and separation; mitigating plasma material interaction (PMI) and minimizing impurity injection; and the degree of axisymmetry required to achieve high I p . Thus, a first of its kind, CHI system is being installed on PEGASUS-III. It utilizes two coaxial, segmented, floating electrodes located entirely within the vacuum vessel in the upper divertor region. The design enables I p scaling studies as the electrode shape, coupled with a new 480 kA/288 kA/244 kA-turn divertor coil triplet, and allows for variation of the flux footprint shape and location simply through manipulation of coil currents. Together, they are projected to allow over 50 mWb connected flux and I p > 300 kA. The segmented electrodes facilitate simple changes to their shape, position, and plasma-facing material. This flexibility may be critical for mitigating PMI or impurity sourcing from the electrodes. Independent current feeds to each segment enable tests of the impact of axisymmetric drive on I p .

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Section: Introductionmentioning

confidence: 99%

A Coaxial Helicity Injection System for Nonsolenoidal Startup Studies on the PEGASUS-III Experiment

Reusch

Raman

Bongard

et al. 2022

IEEE Trans. Plasma Sci.

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“…Upgraded features include: a factor of 4 increase in the toroidal field B T ; new divertor coils; longer pulse length; a new diagnostic neutral beam (DNB); and no central solenoid in the central column. Its nominal machine and plasma parameters are: major radius R 0 ≤ 0.48 m, aspect ratio A ≥ 1.22, plasma current I p ≤ 0.3 MA, B T ≤ 0.6 T, and t pulse ∼ 0.1 s. Sontag et al [5] provides additional information regarding the upgraded facility.…”

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confidence: 99%

Digital Control and Power Systems for the Pegasus-III Experiment

Bongard

Borchardt

Diem

et al. 2022

IEEE Trans. Plasma Sci.

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The PEGASUS-III experiment is a solenoid-free, low aspect ratio spherical tokamak that will serve as a dedicated U.S. platform for comparative nonsolenoidal tokamak plasma startup studies. Approximately 175 megavolt-ampere (MVA) of reconfigured and expanded programmable power systems, 7 MJ of new stored energy, and new digital control and protection systems for the facility are being commissioned to support PEGASUS-III upgrades. These include: increased toroidal field (0.15-0.6 T); new divertor and poloidal field coils; increased pulse length; local and coaxial helicity injectors for solenoid-free plasma initiation; radio frequency (RF) systems for heating and current drive; and a diagnostic neutral beam (DNB). A new realtime digital control system implements 16 proportional-integral differential (PID) feedback controllers with 25 kHz loop rates to control the electromagnets and helicity injectors. The poloidal field coils, helicity injector arc currents, and toroidal field are driven by 36 3.6 MVA (4 kA, 900 V) insulated-gate bipolar transistor (IGBT) buck converters. Helicity injector bias voltage and current will be provided by a set of four 10.8 MVA multilevel buck converters (MLBCs). Each is comprised of an 1800 V integrated gate-commutated thyristor (IGCT) stage and a ±900 V IGBT stage in series, providing controllable I inj ≤ 4 kA at V inj ≤ 2.7 kV. A field programmable gate array (FPGA)-based digital fault protection system multiplexes controller commands to individual power semiconductors in these supplies, monitors their operational status, and executes shutdown sequences within 10 µs of fault detection. An 80 kV, 4 A zero-voltage-switching (ZVS) resonant converter with <1% output ripple is under development for the DNB and is being evaluated as a topology to drive RF sources.

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Effects of Injected Current Streams on MHD Equilibrium Reconstruction of Local Helicity Injection Plasmas in a Spherical Tokamak

Weberski,

Bongard,

Diem

et al. 2024

J Fusion Energ

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Open field line currents are intrinsic to DC helicity injection plasma startup and pose a challenge for inferring the plasma equilibrium with standard reconstruction analysis. Local helicity injection (LHI) is a type of DC helicity injection which uses small, modular current sources to drive force-free current along helical field lines to produce tokamak plasmas. MHD modeling and magnetic measurements during LHI indicate the injected current streams remain coherent as helical structures on the outboard edge of a core toroidal plasma that is tokamak-like in a toroidally averaged sense. To extract core plasma equilibrium properties, external magnetic diagnostics corrected for contributions from the injected current streams are fitted by a standard Grad-Shafranov equilibrium code. An iterative approach for estimating and subtracting the stream contributions from the diagnostic signals is described and applied to a model equilibrium database to reduce systematic errors introduced by the streams. Convergence is usually attained with 2 to 4 iterations, with derived equilibrium parameters matching the prescribed axisymmetric core values to within estimated experimental uncertainties. Accurate recovery of core parameters occurs when the ratio of the net toroidal windup current from the streams to the core plasma current is less than 0.2, which is typically satisfied in most experiments.

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The New PEGASUS-III Experiment

Cited by 4 publications

References 11 publications

A Coaxial Helicity Injection System for Nonsolenoidal Startup Studies on the PEGASUS-III Experiment

A Coaxial Helicity Injection System for Nonsolenoidal Startup Studies on the PEGASUS-III Experiment

Digital Control and Power Systems for the Pegasus-III Experiment

Effects of Injected Current Streams on MHD Equilibrium Reconstruction of Local Helicity Injection Plasmas in a Spherical Tokamak

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