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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Recent Progress in the NSTX/NSTX-U Lithium Program and Prospects for Reactor-Relevant Liquid-Lithium Based Divertor Development

Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. While tungsten has been identified as the most attractive solid divertor material, the NSTX/NSTX-U lithium (Li) program is investigating the viability of liquid lithium (LL) as a potential reactor compatible divertor plasma facing component (PFC) . In the near term, operation in NSTX-U is projected to provide reactor-like divertor heat loads < 40 MW/m^2 for 5 s. During the most recent NSTX campaign, ~ 0.85 kg of Li was evaporated onto the NSTX PFCs where a ~50% reduction in heat load on the Liquid Lithium Divertor (LLD) was observed, attributable to enhanced divertor bolometric radiation. This reduced divertor heat flux through radiation observed in the NSTX LLD experiment is consistent with the results from other lithium experiments and calculations. These results motivate an LL-based closed radiative divertor concept proposed here for NSTX-U and fusion reactors. With an LL coating, the Li is evaporated from the divertor strike point surface due to the intense heat. The evaporated Li is readily ionized by the plasma due to its low ionization energies, and the ionized Li ions can radiate strongly, resulting in a significant reduction in the divertor heat flux. Due to the rapid plasma transport in divertor plasma, the radiation values can be significantly enhanced up to ~ 11 MJ/cc of LL. This radiative process has the desired function of spreading the focused divertor heat load to the entire divertor chamber facilitating the divertor heat removal. The LL divertor surface can also provide a "sacrificial" surface to protect the substrate solid material from transient high heat flux such as the ones caused by the ELMs. The closed radiative LLD concept has the advantages of providing some degree of partition in terms of plasma disruption forces on the LL, Li particle divertor retention, and strong divertor pumping action from the Li-coated divertor chamber wall. By operating at a lower temperature than the first wall, the LLD can serve to purify the entire reactor chamber, as impurities generally migrate toward lower temperature Li-condensed surfaces. To maintain the LL purity, a closed LL loop system with a modest capacity (e.g., ~ 1 Liter/sec for ~ 1% level "impurities") is envisioned for a steady-state 1 GW-electric class fusion power plant.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Snowflake divertor configuration studies in National Spherical Torus Experiment

Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4-6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width {lambda}{sub q} was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5-1 MW/m{sup 2} (from 4-7 MW/m{sup 2} in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, P{sub rad} in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Investigations on the heat flux and impurity for the HL-2M divertor

The controllability of the heat load and impurity in the divertor is very important, which could be one of the critical problems to be solved in order to ensure the success for a steady state tokamak. HL-2M has the advantage of the poloidal field (PF) coils placed inside the demountable toroidal field (TF) coils and close to the main plasma. As a result, it is possible to make highly accurate configuration control of the advanced divertor for HL-2M. The divertor target geometry of HL-2M has been designed to be compatible with different divertor configurations to study the divertor physics and support the high performance plasma operations. In this paper, the heat loads and impurities with different divertor configurations, including the standard X-point divertor, the snowflake-minus divertor and two tripod divertor configurations for HL-2M, are investigated by numerical simulations with the SOLPS5.0 code under the current design of the HL-2M divertor geometry. The plasmas with different conditions, such as the low discharge parameters with ${{I}_{\text{p}}}$ = 0.5 MA at the first stage of HL-2M and the high parameters with ${{I}_{\text{p}}}$ = 2.0 MA during the normal operations, are simulated. The heat load profiles and the impurity distributions are obtained, and the control of the peak heat load and the effect of impurity on the core plasma are discussed. The compatibility of different divertor configurations for HL-2M is also evaluated. It is seen that the excellent compatibility of different divertor configurations with the current divertor geometry has been verified. Here, the results show that the snowflake-minus divertor and the tripod divertor with ${{d}_{x}}=30\,\text{cm}$ present good performance in terms of the heat load profiles and the impurity distributions under different conditions, which may not have a big effect on the core plasma. In addition, it is possible to optimize the distance between the two X-points, ${{d}_{x}}$ , to achieve a better performance in terms of the parameters of discharges.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Neutral leakage, power dissipation and pedestal fueling in open vs closed divertors

The effect of divertor closure on neutral leakage, power dissipation and pedestal fueling is explored using the SOLPS code for different densities including attached and detached divertor conditions for DIII-D open and closed divertors. Here, a metric for divertor closure in terms of the percentage of neutrals escaping the divertor is presented. At the detachment onset the population of neutrals escaping the divertor is 12% for the closed divertor and 35% for the open divertor. This results in a lower contribution of the convective term to the total heat flux and increased power losses in the closed divertor. The comparison between the two geometries reveals that for the same upstream density, the upstream temperature is 23% higher in the closed divertor. The physical mechanisms for detachment at lower upstream density in the closed divertor are also identified with the reduction of the convective term directly resulting in 40% decrease and the increased power losses contributing with a factor of 2. The baffling also lowers the neutral flux refueling the pedestal resulting in further reduction of the core density for achieving detachment. The results introduced here show that the divertor recycling is the primary fueling mechanism in DIII-D. It is shown that fueling occurring in the region between the X-point and the outer midplane is at least four times lower in the closed divertor due the baffling structure which enhances target neutral trapping and consequently reduced upstream leakage. Due to the openness of the inner leg to the common flux and HFS regions both open and closed divertor experience almost identical fueling at the HFS highlighting the need to also consider the inboard divertor for shape optimization.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Modeling of deuterium and carbon radiation transport in MAST-U tokamak advanced divertors

Modest effects of deuterium and carbon radiation opacity in the Super-X and snowflake divertor plasmas are predicted for MAST Upgrade tokamak with core plasma input power 2.5–5 MW and plasma current 1 MA. The radiation transport modeling is based on the SOLPS-EIRENE and UEDGE code divertor plasma predictions. Two radiation transport models are used: one is based on a full radiation transport equation implemented in the radiation transport and collisional-radiative code CRETIN (without feedback on the background plasma), and another is an internal self-consistent UEDGE model with ionization, recombination, and heating rates corrected for Ly α line trapping based on the escape probability model implemented in CRETIN. In MAST-U, the Super-X and snowflake divertor plasmas are predicted to reach detached regimes at lower upstream densities than the standard divertor, and the conclusion still holds with radiation transport effects included. At neutral densities m -3 , modest Ly α deuterium line trapping with optical depths 10–15 is predicted in the Super-X divertor. Divertor plasmas are optically thin to other Lyman and Balmer lines, as well as to strong C III and C IV lines that are responsible for most of divertor radiated power. Insignificant changes (within a few percent) to divertor deuterium ionization and recombination rates are found. Radiation fluxes on outer divertor target are modified within a factor of 2–3 when the radiation transport is accounted for, and a similar variation is found due to the line shape models that define the absorption and emission line profiles in the radiation transport modeling. The predicted Lyman and Balmer spectral intensities are significantly modified due to radiation trapping. A measurement of divertor radiation transport effects is discussed using the Ly β /Ba α line ratio. In the snowflake divertor configuration, divertor plasmas are found to be optically thin to Lyman series lines within a large range of parameter variations that include upstream density, divertor transport coefficients, and magnetic configurations. Finally, modest radiation transport effects are only found in a few cases with strongest divertor transport and magnetic configurations closest to the ideal snowflake configuration, however, the plasma background models that were used are yet to be validated with an experiment.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2011

Snowflake divertor configuration studies for NSTX-Upgrade

Snowflake divertor experiments in NSTX provide basis for PMI development toward NSTX-Upgrade. Snowflake configuration formation was followed by radiative detachment. Significant reduction of steady-state divertor heat flux observed in snowflake divertor. Impulsive heat loads due to Type I ELMs are partially mitigated in snowflake divertor. Magnetic control of snowflake divertor configuration is being developed. Plasma material interface development is critical for NSTX-U success. Four divertor coils should enable flexibility in boundary shaping and control in NSTX-U. Snowflake divertor experiments in NSTX provide good basis for PMI development in NSTX-Upgrade. FY 2009-2010 snowflake divertor experiments in NSTX: (1) Helped understand control of magnetic properties; (2) Core H-mode confinement unchanged; (3) Core and edge carbon concentration reduced; and (4) Divertor heat flux significantly reduced - (a) Steady-state reduction due to geometry and radiative detachment, (b) Encouraging results for transient heat flux handling, (c) Combined with impurity-seeded radiative divertor. Outlook for snowflake divertor in NSTX-Upgrade: (1) 2D fluid modeling of snowflake divertor properties scaling - (a) Edge and divertor transport, radiation, detachment threshold, (b) Compatibility with cryo-pump and lithium conditioning; (2) Magnetic control development; and (3) PFC development - PFC alignment and PFC material choice.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

High performance double-null plasmas under radiating divertor and mantle scenarios on DIII-D

Experiments to combine a radiating divertor or mantle with high power, high βN plasma operation have successfully reduced the divertor heat flux by 40% but have also encountered several challenging problems. For example, injecting either neon or argon seed impurities have led both to significant fuel dilution of the core plasma and to the emergence of harmful tearing mode activity which compromised plasma beta and energy confinement time. Increased divertor closure ended in more effective control over the injected seed impurity inventory, although it was also observed that small radial variations in the placement of the outer separatrix strike point within the DIII-D ‘closed’ divertor could lead to clear differences in the impurity build-up inside the core plasma. Active particle pumping of argon seed impurities through the divertor leg on the high-field side of a double-null divertor configuration showed little effect on seed impurity inventory and only modest control over the deuterium fuel inventory. Normally, 80% or greater of injected argon was removed by the corresponding cryo-pump on the low-field (outboard) side of that divertor, with the remaining argon being removed by the outboard cryo-pump from the opposite divertor. For single-null divertor cases, the contribution of the inner (high-field) cryo-pump to impurity and main-ion control appears to be complicated by the levels of divertor recycling and degree of divertor detachment. In general, divertor design that impedes the escape of seed impurities, plasma shaping that allows a higher stability limit, and electron cyclotron deposition profiles that prevent tearing mode onset and impede impurity accumulation in the core are under consideration for addressing the issues presented in our work.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

SOLPS-ITER modeling of a dedicated divertor for negative triangularity operation on DIII-D

The design of a new dedicated divertor for negative triangularity (NT) operation on DIII-D with neutral baffles and pumping is informed by SOLPS-ITER transport modeling. This dedicated NT divertor is the latest step in a progression of NT shapes with various divertor characteristics explored on DIII-D, including NT shapes at reduced triangularity and a campaign with stronger shaping that included new armored components on the outboard side. SOLPS simulations played a key role in these divertor designs. Interpretive simulations, using cross-field diffusivities constrained by experimental data in the NT Shelf shape were used to inform the design of the 2023 armor campaign components. A similar procedure used armor campaign data to predict conditions for the dedicated NT divertor. The predictive simulations were used to assess the divertor fluxes, detachment threshold, pumped flux, and neutral leakage. For the dedicated NT divertor, SOLPS simulations and two-point-modeling were used to show the relative impact of magnetic topology (mainly longer connection length) and divertor closure on the divertor conditions relative to the armor campaign. It is predicted that the dedicated NT divertor reaches detachment (measured by target ion flux rollover) at a lower upstream density (≈(1.75−−2.4)×1⁢019m−3) as compared to the armor campaign shape. For the preliminary design geometry, divertor closure reduces the neutral leakage by ≈10%. Parametric optimization indicating further ≈20%–60% improvement in the leakage flux and recycled flux crossing the pump entrance is possible for relatively minor changes to the divertor and baffle layout.

Lore, Jeremy [ORNL] (ORCID:000000029192465X)↗
DOE OSTI2017

A review of radiative detachment studies in tokamak advanced magnetic divertor configurations

The present vision for a plasma–material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertor configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). As a result, this paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Performance assessment of long-legged tightly-baffled divertor geometries in the ARC reactor concept

Extremely intense power exhaust channels are projected for tokamak-based fusion power reactors; a means to handle them remains to be demonstrated. Advanced divertor configurations have been proposed as potential solutions. Recent modelling of tightly baffled, long-legged divertor geometries for the divertor test tokamak concept, ADX, has shown that these concepts may access passively stable, fully detached regimes over a broad range of parameters. The question remains as to how such divertors may perform in a reactor setting. To explore this, numerical simulations are performed with UEDGE for the long-legged divertor geometry proposed for the ARC pilot plant conceptual design - a device with projected heat flux power width (λq||) of 0.4 mm and power exhaust of 93 MW - first for a simplified Super-X divertor configuration (SXD) and then for the actual X-point target divertor (XPTD) being proposed. It is found that the SXD, combined with 0.5% fixed-fraction neon impurity concentration, can produce passively stable, detached divertor regimes for power exhausts in the range of 80-108 MW - fully accommodating ARC's power exhaust. The XPTD configuration is found to reduce the strike-point temperature by a factor of ~10 compared to the SXD for small separations (~1.4λq||) between main and divertor X-point magnetic flux surfaces. Even greater potential reductions are identified for reducing separations to ~1λq|| or less. The power handling response is found to be insensitive to the level of cross-field convective or diffusive transport assumed in the divertor leg. By raising the separatrix density by a factor of 1.5, stable fully detached divertor solutions are obtained that fully accommodate the ARC exhaust power without impurity seeding. To our knowledge, this is the first time an impurity-free divertor power handling scenario has been obtained in edge modelling for a tokamak fusion power reactor with λq|| of 0.4 mm.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗