Tungsten erosion and divertor leakage from the DIII-D SAS-VW tungsten-coated divertor in experiments with neon gas seeding

Parsons, Matthew S.; Mateja, Jeremy D.; Messer, Seth H.; Abrams, Tyler; Allain, Jean Paul; Bortolon, Alessandro; Byrne, Patrick; Donovan, David C.; Effenberg, Florian; Herfindal, Jeffrey L.; Laggner, Florian; Odstrcil, Tomas; Ren, Jun; Rudakov, Dmitry L.; Sinclair, Gregory; Wilcox, Robert S.

doi:10.1016/j.nme.2023.101520

Cited by 3 publications

(3 citation statements)

References 56 publications

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“…Selected discharge parameters from discharge 190 783, where neon was injected at multiple rates from the SOL crown, are shown in figure 2. In the discharge with constant conditions, graphite Collector Probes were exposed to the plasma at both the outer midplane and at the lower shelf in order to measure the tungsten divertor leakage, as discussed in detail elsewhere [17]. At each location, the discharge number with solid font was a scan of injection rates while the discharge with outlined font was performed at one constant injection rate.…”

Section: Methodsmentioning

confidence: 99%

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Interpretive modeling of tungsten divertor leakage during experiments with neon gas seeding

Parsons,

Sinclair,

Abrams

et al. 2024

Nucl. Fusion

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Many existing and future tokamaks with tungsten divertors operate, or will operate, with low-Z impurity seeding, but the direct effect of these seeded impurities on tungsten Scrape-off-Layer (SOL) transport has not been explored in detail. This paper reports on a DIII-D experiment designed to test how tungsten divertor leakage from the Small- Angle Slot V-Shaped, tungsten-coated divertor is impacted by neon seeding at a variety of injection rates and poloidal injection locations. Measurements from the experiment show an inverse relationship between the neon injection rate and the tungsten core penetration factor. Interpretive modeling is performed with a combination of the SOLPS-ITER and DIVIMP codes to assess the underlying tungsten behavior. The modeling results show that the reduction in tungsten divertor leakage is driven by both an increase in the divertor collisionality as well as a reduction in the ion temperature gradient near the divertor target. Collisions between low-Z impurities and tungsten impurities are found to have a significant impact on the tungsten SOL transport, such that ignoring the low-Z impurity collisional effects on the tungsten transport can result in an overestimate of the divertor leakage by an order-of-magnitude. Given the importance of these localized interactions, neon seeding from the closed, slot-like divertor has a clear advantage in being able to reduce tungsten divertor leakage without the high levels of neon core contamination that occur when seeding from other poloidal locations.

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Section: Methodsmentioning

confidence: 99%

“…The 4 neon injection locations are indicated, along with the locations of the cryopumps.At each location, the discharge number with solid font was a scan of injection rates while the discharge with outlined font was performed at one constant injection rate. Reproduced from[17]. CC BY 4.0.…”

mentioning

confidence: 99%

Interpretive modeling of tungsten divertor leakage during experiments with neon gas seeding

Parsons,

Sinclair,

Abrams

et al. 2024

Nucl. Fusion

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“…Hence, tungsten's presence was not included in the modelling, and the slot is denoted as 'SAS-V' in the modelling but 'SAS-VW' in the experiment. For an in-depth analysis of the tungsten measurements from the same SAS-VW campaign, refer to [16][17][18].…”

Section: Introductionmentioning

confidence: 99%

Experiments on plasma detachment in a V-shaped slot divertor in the DIII-D tokamak

Maurizio,

Thomas,

et al. 2024

Nucl. Fusion

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Experiments in DIII-D demonstrate that the upstream plasma density to detach an un-pumped slot divertor is similar for a V-shaped and a flat-end slot, despite significantly higher neutral pressure in the V-shaped slot and in contrast to SOLPS-ITER predictions. The detachment threshold can be reduced by using in-slot instead of main-chamber gas fuelling or by placing the strike point on the inner slanted slot baffle instead of the slot end, as described by simulations with full drift physics. When increasing the plasma line-averaged density (without extrinsic impurities), the transition to detachment in DIII-D slot divertor is sharp and requires a high value of plasma density with the ion B → × ∇ B drift into the slot, whereas it is smooth and requires a lower value of plasma density with the opposite drift direction, in accord with detachment experiments in the DIII-D open lower divertor. Unique experiments on DIII-D and comparison to advanced simulations expand the scientific understanding of slot-shaped divertors, considered highly desirable for next step fusion devices.

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Managing the heat: In-Vessel Components

Cane,

Barth,

Farrington

et al. 2024

Phil. Trans. R. Soc. A.

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The Spherical Tokamak for Energy Production (STEP) programme aims to deliver a first-of-a-kind fusion prototype powerplant (SPP). The SPP plasma places extreme heat, particle and structural loads onto the plasma-facing components (PFCs) of the divertor, limiters and inboard and outboard sections of the first wall. The PFCs must manage the heat and particle loads and wider powerplant requirements relating to safety, net power generation, tritium breeding and plant availability. To enable STEP PFC concepts to be identified that satisfy these wide-ranging requirements, an iterative design (‘Decide & Iterate’) methodology has been used to synchronize a prioritized set of decisions, within the fast-paced, iterative, whole plant concept design schedule. This paper details the ‘Decide and Iterate’ methodology and explains how it has enabled the identification of the SPP PFC concepts. These include innovative PFC solutions such as a helium-cooled discrete and panel limiter design to increase tritium breeding while providing sufficient coverage and enabling individual limiter replacement; the integration of the outboard first wall with the breeding zone to enhance fuel self-sufficiency and power generation; and the use of heavy water (D 2 O) within the inboard first wall and divertor PFCs to increase tritium breeding within the outboard breeding zone. This article is part of the theme issue ‘Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)’.

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Tungsten erosion and divertor leakage from the DIII-D SAS-VW tungsten-coated divertor in experiments with neon gas seeding

Cited by 3 publications

References 56 publications

Interpretive modeling of tungsten divertor leakage during experiments with neon gas seeding

Interpretive modeling of tungsten divertor leakage during experiments with neon gas seeding

Experiments on plasma detachment in a V-shaped slot divertor in the DIII-D tokamak

Managing the heat: In-Vessel Components

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