Active Divertor Heat Flux Control using Impurity Powder Dropper

Ono, M.; Raman, R.; Maingi, R.; Kaye, S.; Sanchez-Villar, A.

doi:10.1007/s10894-025-00513-3

Active Divertor Heat Flux Control using Impurity Powder Dropper

Journal Article · Mon Oct 06 00:00:00 EDT 2025 · Journal of Fusion Energy (Online)

DOI:https://doi.org/10.1007/s10894-025-00513-3· OSTI ID:2999372

^[1]; ^[2]; ^[1]; ^[1]; ^[1]

Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States)
Univ. of Washington, Seattle, WA (United States)

Divertor plasma-facing components (PFCs) in a tokamak are typically designed to withstand average steady-state heat loads of about 5–10 MW/m², a limit that applies to both solid and liquid lithium (LL) PFCs. Exceeding these design values can result in surface damage to tungsten PFCs or excessive lithium (Li) evaporation in liquid lithium divertor (LLD) PFCs. Since exceeding the divertor heat load limits has serious consequences, it is therefore prudent to develop a tool to reduce the divertor heat load and bring the heat load to within the design limit without affecting the plasma performance. Active low Z impurity injection such as Li has been suggested as a potential solution to mitigate excess heat flux as suggested previously, given that non-coronal radiation can be quite large ~ 20–30 MJ per mole of injected Li. Li is considered desirable for reducing the edge neutral recycling helping to improve plasma energy confinement. In this paper, we model the Impurity Power Dropper (IPD) to investigate its potential of divertor heat flux control. The IPD is typically located at the top of the tokamak device and uses a vertical drift tube of a few meters. In the 2 m drift tube case, the IPD powder is accelerated to ~ 6 m/sec before reaching the plasma with the upper divertor configuration, matching the condition for the in-board side pellet injection case. By modeling the IPD geometry we determined the IPD powder deposition profile, and thus the non-coronal radiation and ionization profiles in time as well. From the enhanced radiation power loss, it is therefore possible to reduce the divertor heat load using the divertor simulation code. In conclusion, the IPD divertor heat flux control can be tested in the facilities with IPD including ST-40, DIII-D, EAST, WEST and NSTX-U.

View Accepted Manuscript (DOE)

Research Organization:: Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States)

Sponsoring Organization:: USDOE

Grant/Contract Number:: AC02-09CH11466; SC0006757

OSTI ID:: 2999372

Journal Information:: Journal of Fusion Energy (Online), Journal Name: Journal of Fusion Energy (Online) Journal Issue: 2 Vol. 44; ISSN 1572-9591

Publisher:: SpringerCopyright Statement

Country of Publication:: United States

Language:: English

References (23)

Radiative Cooling Rates for Low-Z Impurities in Non-coronal Equilibrium State Mavrin, A. A. Journal of Fusion Energy, Vol. 36, Issue 4-5 https://doi.org/10.1007/s10894-017-0136-z	journal	July 2017
Active Radiative Liquid Lithium Divertor for Handling Transient High Heat Flux Events Ono, M.; Raman, R. Journal of Fusion Energy, Vol. 39, Issue 6 https://doi.org/10.1007/s10894-020-00253-6	journal	September 2020
Active Lithium Injection for a Real Time Control of the Divertor Heat Flux for Fusion Devices Ono, M.; Raman, R. Journal of Fusion Energy, Vol. 42, Issue 2 https://doi.org/10.1007/s10894-023-00387-3	journal	October 2023
Recent improvements of the helium-cooled W-based divertor for fusion power plants Wang, X. R.; Malang, S.; Tillack, M. S. Fusion Engineering and Design, Vol. 87, Issue 5-6 https://doi.org/10.1016/j.fusengdes.2012.02.012	journal	August 2012
Manufacturing and testing of a HETS module for DEMO divertor Visca, Eliseo; Agostini, Pietro; Crescenzi, Fabio Fusion Engineering and Design, Vol. 87, Issue 7-8 https://doi.org/10.1016/j.fusengdes.2012.02.070	journal	August 2012
Active radiative liquid lithium divertor concept Ono, M.; Jaworski, M. A.; Kaita, R. Fusion Engineering and Design, Vol. 89, Issue 12 https://doi.org/10.1016/j.fusengdes.2014.05.008	journal	December 2014
Real-time reduction of tungsten impurity influx using lithium powder injection in EAST Xu, W.; Hu, J. S.; Maingi, R. Fusion Engineering and Design, Vol. 137 https://doi.org/10.1016/j.fusengdes.2018.09.009	journal	December 2018
Making tungsten work – ICFRM-14 session T26 paper 501 Nygren et al. making tungsten work Nygren, R. E.; Raffray, R.; Whyte, D. Journal of Nuclear Materials, Vol. 417, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2010.12.289	journal	October 2011
Recent lithium experiments in tokamak T-11M Mirnov, S. V.; Belov, A. M.; Djigailo, N. T. Journal of Nuclear Materials, Vol. 438 https://doi.org/10.1016/j.jnucmat.2013.01.032	journal	July 2013
Physics basis for the first ITER tungsten divertor Pitts, R. A.; Bonnin, X.; Escourbiac, F. Nuclear Materials and Energy, Vol. 20 https://doi.org/10.1016/j.nme.2019.100696	journal	August 2019
Impurity transport in edge plasmas with application to liquid walls Rognlien, T. D.; Rensink, M. E. Physics of Plasmas, Vol. 9, Issue 5 https://doi.org/10.1063/1.1461384	journal	May 2002
Edge-localized-modes in tokamaks Leonard, A. W. Physics of Plasmas, Vol. 21, Issue 9 https://doi.org/10.1063/1.4894742	journal	September 2014
Erratum: “Charge breeding of radioactive isotopes at the CARIBU facility with an electron beam ion source” [Rev. Sci. Instrum. 89, 052402 (2018)] Vondrasek, R. C.; Dickerson, C. A.; Hendricks, M. Review of Scientific Instruments, Vol. 89, Issue 10 https://doi.org/10.1063/1.5063798	journal	October 2018
Constructing a new predictive scaling formula for ITER's divertor heat-load width informed by a simulation-anchored machine learning Chang, C. S.; Ku, S.; Hager, R. Physics of Plasmas, Vol. 28, Issue 2 https://doi.org/10.1063/5.0027637	journal	February 2021
A model for the ablation rate of a solid hydrogen pellet in a plasma Parks, P. B.; Turnbull, R. J.; Foster, C. A. Nuclear Fusion, Vol. 17, Issue 3 https://doi.org/10.1088/0029-5515/17/3/013	journal	June 1977
A magnetohydrodynamic simulation of pellet ablation in the electrostatic approximation Samulyak, Roman; Lu, Tianshi; Parks, Paul Nuclear Fusion, Vol. 47, Issue 2 https://doi.org/10.1088/0029-5515/47/2/005	journal	January 2007
Compact DEMO, SlimCS: design progress and issues Tobita, K.; Nishio, S.; Enoeda, M. Nuclear Fusion, Vol. 49, Issue 7 https://doi.org/10.1088/0029-5515/49/7/075029	journal	July 2009
Scaling of the tokamak near the scrape-off layer H-mode power width and implications for ITER Eich, T.; Leonard, A. W.; Pitts, R. A. Nuclear Fusion, Vol. 53, Issue 9 https://doi.org/10.1088/0029-5515/53/9/093031	journal	August 2013
Enhanced H-mode pedestals with lithium injection in DIII-D Osborne, T. H.; Jackson, G. L.; Yan, Z. Nuclear Fusion, Vol. 55, Issue 6 https://doi.org/10.1088/0029-5515/55/6/063018	journal	May 2015
The ITER design Aymar, R.; Barabaschi, P.; Shimomura, Y. Plasma Physics and Controlled Fusion, Vol. 44, Issue 5 https://doi.org/10.1088/0741-3335/44/5/304	journal	April 2002
Lithium divertor concept and results of supporting experiments Evtikhin, V. A.; Lyublinski, I. E.; Vertkov, A. V. Plasma Physics and Controlled Fusion, Vol. 44, Issue 6 https://doi.org/10.1088/0741-3335/44/6/322	journal	May 2002
ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor Maingi, R.; Hu, J. S.; Sun, Z. Nuclear Fusion, Vol. 58, Issue 2 https://doi.org/10.1088/1741-4326/aa9e3f	journal	January 2018
Massive Gas Injection Valve Development for NSTX-U Raman, R.; Plunkett, G. J.; Lay, W. -S. IEEE Transactions on Plasma Science, Vol. 44, Issue 9 https://doi.org/10.1109/TPS.2016.2565658	journal	September 2016

Similar Records

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

Conference · Sat Oct 27 00:00:00 EDT 2012 · IAEA · OSTI ID:1056358

Full-torus impurity transport simulation for optimizing plasma discharge operation using a multi-species impurity powder dropper in the large helical device

Journal Article · Wed Dec 11 23:00:00 EST 2019 · Contributions to Plasma Physics · OSTI ID:1644264

Related Subjects

Active lithium injection
Divertor
High Temperature Plasma
Impurity powder dropper
Liquid lithium
Low Temperature Plasma
Magnetic and inertial plasma confinement
Nuclear Fusion
Plasma processing
Tokamaks
Transient heat flux

Active Divertor Heat Flux Control using Impurity Powder Dropper

Citation Formats

References (23)

Similar Records

Related Subjects