Bulletin of the American Physical Society
64th Annual Meeting of the APS Division of Plasma Physics
Volume 67, Number 15
Monday–Friday, October 17–21, 2022; Spokane, Washington
Session YM10: Mini-Conference: The Integrated Tokamak Exhaust and Performance Gap IVLive Streamed
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Chair: Brian Grierson, General Atomics Room: 206 CD |
Friday, October 21, 2022 9:30AM - 9:35AM |
YM10.00001: Introductory Remarks
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Friday, October 21, 2022 9:35AM - 9:55AM |
YM10.00002: Using Pedestal Physics to Close the Integrated Tokamak Performance and Exhaust (ITEP) Gap Philip B Snyder, John Canik, Jerry W Hughes, Matthias Knolker, Orso Meneghini, Thomas H Osborne, Jin Myung Park, Wayne M Solomon, Robert S Wilcox, Theresa M Wilks, Howard R Wilson The edge transport barrier, or "pedestal," provides the interface between the hot core fusion plasma, and the much cooler boundary plasma which exhausts heat and particles from the system. To close the ITEP gap, an order of magnitude improvement in the product of core fusion performance (measured by the product of pressure and bootstrap fraction) and parallel heat exhaust capability (measured by PB/R) is required, relative to both existing devices and ITER. Here we employ an updated version of the EPED pedestal model, both independently and coupled to core transport (TGLF) and boundary physics (SOLPS) models, to develop and optimize regimes which potentially close the ITEP gap. Strong shaping, moderate aspect ratio (R/a~2.3-2.7), and high field (B>~6T) facilitate operation with a high pressure pedestal limited by current-driven kink/peeling modes ("peeling limited," including the "Super H-mode" regime) even at relatively high density. In this peeling-limited regime, the pedestal is predicted not to be degraded by high separatrix density and pressure, facilitating compatibility with a dense radiative divertor plasma. Promising regimes are identified both for a demonstration device and a compact fusion pilot plant. |
Friday, October 21, 2022 9:55AM - 10:15AM |
YM10.00003: Narrowing the ITEP Pedestal Physics GAP with Strong Plasma Shaping at High Plasma Current on DIII-D Tom Osborne Combining high core performance with a dissipative divertor in an H-mode based fusion power plant (FPP) requires high pedestal pressure, pPED, at high density, nePED. These pedestals will be in a regime of high opacity to neutrals entering from the scrape off layer but low collisionality: conditions that are difficult to achieve simultaneously in present experiments. DIII-D will exploit the Super-H-mode regime at high plasma current, IP, combined with strong plasma shaping to build a better understanding of the physics setting pedestal structure with more FPP relevant neutral opacity and collisionality. Modification of the divertor geometry is planned to allow density control in these high triangularity, high elongation, large minor radius discharges as well as an increase in toroidal field to extend Super-H-mode access to higher IP. In the Super-H-mode regime pPED increases with nePED, IP and triangularity, while Super-H access will be lost if the safety factor, q, is too low. High elongation and the optimal squareness are required also to relax the level of nePED and TPED control needed to enter the regime. Edge particle source widths down to 20% of the nePED width at ν*e ~ 0.5 with nePED~1.6x1020/m3, nePED/nG= 0.8 at IP=2.3MA, q95=4 may be achievable. |
Friday, October 21, 2022 10:15AM - 10:35AM |
YM10.00004: Negative triangularity plasmas in DIII-D and their potential for core-edge integration Filippo Scotti, Max Austin, Alessandro Marinoni, Carlos A Paz-Soldan, Kathreen E Thome In DIII-D, discharges with negative triangularity (NegD) shaping (δu~-0.4) display H-mode-like confinement (H98,y2~1, βN~3) without ELMs or a pedestal with input power PINJ up to 14MW (~4xPLH). Improved confinement with respect to positive triangularity L-mode is due to reduced turbulent transport while ballooning mode destabilization prevents H-mode access. Poor core impurity confinement is observed with confinement times τZ<100ms. NegD discharges display L-mode like edge ne gradients and edge Te gradients intermediate between L and H-mode, with SOL power widths (λq) in NegD L-mode discharges up to 50% wider than in H-mode plasmas. While dissipative regimes were not explored experimentally, fluid simulations (SOLPS, UEDGE) with PSOL~5 MW identified a window of detachment access without impurity seeding. The wider λq and weak density peaking are expected to be beneficial for detachment access at low average ne while improving detachment controllability. L-mode operation at high PINJ and short τZ can enable scenarios with high core radiation fraction, easing requirements on divertor dissipation, with the potential to provide an integrated core-edge solution for a fusion power plant. |
Friday, October 21, 2022 10:35AM - 10:55AM |
YM10.00005: Utilization of main-chamber limiters for mitigation of steady-state first-wall heat and particle fluxes at reactor scale Jacob H Nichols, Peter C Stangeby, E.A. Unterberg, Shawn A Zamperini, Tyler Abrams Plasma contact at the first wall represents a key point of tension for the integrated design of a self-sufficient tokamak reactor, as the wall cladding must be very thin (< 5 mm) to allow for adequate tritium breeding, but must also withstand significant steady-state heat fluxes and sputter erosion. This issue is exacerbated by the experimental observation that main-chamber ion fluxes increase when operating the divertor at high density. Introducing main-chamber limiters may help alleviate this tension: these limiters, made of more robust materials than the rest of the first wall, would intercept ionic plasma fluxes in the far scrape-off layer before these fluxes can reach the fragile recessed breeding wall. However, careful limiter optimization is needed, as installing limiters over too large an area reduces the overall tritium breeding capability, and installing them too deep into the plasma leads to excessive limiter damage and plasma degradation. Heat flux modeling is presented for a variety of main-chamber limiter designs, using the ARIES-ACT2 reactor design as a starting point. It is found that toroidally-continuous main-chamber limiters are especially promising, as they protect the recessed wall without excessive heat flux peaking on the limiters themselves. Toroidal limiters have the additional benefit of reducing connection lengths in front of the breeding wall, which is theorized to reduce the cross-field propagation of plasma filaments and thus further reduce ion fluxes to recessed surfaces. |
Friday, October 21, 2022 10:55AM - 11:50AM |
YM10.00006: Panel Discussions
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