Conceptual design of a pre-loaded liquid lithium divertor target for NSTX-U

Rindt, P.; Lopes Cardozo, N. J.; van Dommelen, J. A. W.; Kaita, R.; Jaworski, M. A.

doi:10.1016/j.fusengdes.2016.08.020

Title: Conceptual design of a pre-loaded liquid lithium divertor target for NSTX-U

Journal Article · Sat Sep 03 00:00:00 EDT 2016 · Fusion Engineering and Design

DOI:https://doi.org/10.1016/j.fusengdes.2016.08.020· OSTI ID:1335696

Rindt, P. ^[1]; Lopes Cardozo, N. J. ^[1]; van Dommelen, J. A. W. ^[1];

^[2]; Jaworski, M. A. ^[2]

Eindhoven Univ. of Technology, Eindhoven (The Netherlands)
Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

In this study, a conceptual design for a pre-filled liquid lithium divertor target for the National Spherical Torus Experiment Upgrade (NSTX-U) is presented. The design is aimed at facilitating experiments with high lithium flux from the plasma facing components (PFCs) in NSTX-U and investigating the potential of capillary based liquid lithium components. In the design, lithium is supplied from a reservoir in the PFC to the plasma facing surface via capillary action in a wicking structure. This working principle is also demonstrated experimentally. Next, a titanium zirconium molybdenum (TZM) prototype design is presented, required to withstand a steady state heat flux peaking at 10 MW m^–2 for 5 s and edge localized modes depositing (130 kJ in 2 ms at 10 Hz). The main challenge is to sufficiently reduce the thermal stresses. This is achieved by dividing the surface into brushes and filling the slots in between with liquid lithium. The principle of using this liquid “interlayer” allows for thermal expansion and simultaneously heat conduction, and could be used to significantly reduce the demands to solids in future PFCs. Lithium flow to the surface is analyzed using a novel analytical model, ideally suited for design purposes. Thermal stresses in the PFC are analyzed using the finite element method. As a result, the requirements are met, and thus a prototype will be manufactured for physical testing.

View Accepted Manuscript (DOE)

Cite

Export

Save

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

Sponsoring Organization:: USDOE

OSTI ID:: 1335696

Journal Information:: Fusion Engineering and Design, Vol. 112, Issue C; ISSN 0920-3796

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

Citation Metrics:

Cited by: 11 works

Citation information provided by
Web of Science

References (27)

Recent analysis of key plasma wall interactions issues for ITER Roth, Joachim; Tsitrone, E.; Loarte, A. Journal of Nuclear Materials, Vol. 390-391 https://doi.org/10.1016/j.jnucmat.2009.01.037	journal	June 2009
Experiments with lithium limiter on T-11M tokamak and applications of the lithium capillary-pore system in future fusion reactor devices Mirnov, S. V.; Azizov, E. A.; Evtikhin, V. A. Plasma Physics and Controlled Fusion, Vol. 48, Issue 6 https://doi.org/10.1088/0741-3335/48/6/009	journal	May 2006
Self-Regulated Plasma Heat Flux Mitigation Due to Liquid Sn Vapor Shielding van Eden, G. G.; Morgan, T. W.; Aussems, D. U. B. Physical Review Letters, Vol. 116, Issue 13 https://doi.org/10.1103/PhysRevLett.116.135002	journal	April 2016
Surface chemistry analysis of lithium conditioned NSTX graphite tiles correlated to plasma performance Taylor, C. N.; Luitjohan, K. E.; Heim, B. Fusion Engineering and Design, Vol. 88, Issue 12 https://doi.org/10.1016/j.fusengdes.2013.09.007	journal	December 2013
Testing of liquid lithium limiters in CDX-U Majeski, R.; Kaita, R.; Boaz, M. Fusion Engineering and Design, Vol. 72, Issue 1-3 https://doi.org/10.1016/j.fusengdes.2004.07.002	journal	November 2004
FTU results with a liquid lithium limiter Mazzitelli, G.; Apicella, M. L.; Frigione, D. Nuclear Fusion, Vol. 51, Issue 7 https://doi.org/10.1088/0029-5515/51/7/073006	journal	May 2011
Liquid lithium divertor characteristics and plasma–material interactions in NSTX high-performance plasmas Jaworski, M. A.; Abrams, T.; Allain, J. P. Nuclear Fusion, Vol. 53, Issue 8 https://doi.org/10.1088/0029-5515/53/8/083032	journal	July 2013
Calculation and experimental investigation of fusion reactor divertor plate and first wall protection by capillary-pore systems with lithium Evtikhin, V. A.; Lyublinski, I. E.; Vertkov, A. V. Journal of Nuclear Materials, Vol. 271-272 https://doi.org/10.1016/S0022-3115(98)00793-4	journal	May 1999
First results of flowing liquid lithium limiter in HT-7 Ren, J.; Hu, J. S.; Zuo, G. Z. Physica Scripta, Vol. T159 https://doi.org/10.1088/0031-8949/2014/T159/014033	journal	April 2014
Erosion of lithium coatings on TZM molybdenum and graphite during high-flux plasma bombardment Abrams, T.; Jaworski, M. A.; Kaita, R. Fusion Engineering and Design, Vol. 89, Issue 12 https://doi.org/10.1016/j.fusengdes.2014.06.005	journal	December 2014
Overview of the physics and engineering design of NSTX upgrade Menard, J. E.; Gerhardt, S.; Bell, M. Nuclear Fusion, Vol. 52, Issue 8 https://doi.org/10.1088/0029-5515/52/8/083015	journal	July 2012
Dependence of divertor heat flux widths on heating power, flux expansion, and plasma current in the NSTX Gray, T. K.; Maingi, R.; Soukhanovskii, V. A. Journal of Nuclear Materials, Vol. 415, Issue 1 https://doi.org/10.1016/j.jnucmat.2011.01.029	journal	August 2011
Capillary wicking of liquid lithium on laser textured surfaces for plasma facing components Lin, T. F.; Palmer, T. A.; Meinert, K. C. Journal of Nuclear Materials, Vol. 433, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2012.08.011	journal	February 2013
Recrystallization kinetics of warm-rolled tungsten in the temperature range 1150–1350 °C Alfonso, A.; Juul Jensen, D.; Luo, G. -N. Journal of Nuclear Materials, Vol. 455, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2014.08.037	journal	December 2014
La2O3 effects on TZM alloy recovery, recrystallization and mechanical properties Wang, Kuai-She; Tan, Jiang-Fei; Hu, Ping Materials Science and Engineering: A, Vol. 636 https://doi.org/10.1016/j.msea.2015.03.114	journal	June 2015
Permeability of open-pore microcellular materials Despois, J.; Mortensen, A. Acta Materialia, Vol. 53, Issue 5 https://doi.org/10.1016/j.actamat.2004.11.031	journal	March 2005
Modeling the reduction of gross lithium erosion observed under high-flux deuterium bombardment Abrams, T.; Jaworski, M. A.; Kaita, R. Journal of Nuclear Materials, Vol. 463 https://doi.org/10.1016/j.jnucmat.2014.11.056	journal	August 2015
High heat flux capabilities of the Magnum-PSI linear plasma device De Temmerman, G.; van den Berg, M. A.; Scholten, J. Fusion Engineering and Design, Vol. 88, Issue 6-8 https://doi.org/10.1016/j.fusengdes.2013.05.047	journal	October 2013
Improvement in Recrystallization Temperature and Mechanical Properties of a Commercial TZM Alloy through Microstructure Control by Multi-Step Internal Nitriding Nagae, Masahiro; Yoshio, Tetsuo; Takada, Jun MATERIALS TRANSACTIONS, Vol. 46, Issue 10 https://doi.org/10.2320/matertrans.46.2129	journal	January 2005
Dynamic Young’s moduli of tungsten and tantalum at high temperature and stress Škoro, G. P.; Bennett, J. R. J.; Edgecock, T. R. Journal of Nuclear Materials, Vol. 409, Issue 1 https://doi.org/10.1016/j.jnucmat.2010.12.222	journal	February 2011
Thermal Conductivity of the Elements Ho, C. Y.; Powell, R. W.; Liley, P. E. Journal of Physical and Chemical Reference Data, Vol. 1, Issue 2 https://doi.org/10.1063/1.3253100	journal	April 1972
Heat Capacity of Reference Materials: Cu and W White, G. K.; Collocott, S. J. Journal of Physical and Chemical Reference Data, Vol. 13, Issue 4 https://doi.org/10.1063/1.555728	journal	October 1984
Wetting properties of liquid lithium on select fusion relevant surfaces Fiflis, P.; Press, A.; Xu, W. Fusion Engineering and Design, Vol. 89, Issue 12 https://doi.org/10.1016/j.fusengdes.2014.03.060	journal	December 2014
Design Models and Experimental Data for Conceptual Design of a Pre-Loaded Liquid Lithium Divertor Target for NSTX-U Rindt, Peter Eindhoven University of Technology https://doi.org/10.4121/uuid:17debbe0-9a08-4807-b957-c504bf1aaf72	dataset	January 2016
Development of a liquid-metal fusion reactor divertor with a capillary-pore system Golubchikov, L. G.; Evtikhin, V. A.; Lyublinski, I. E. Journal of Nuclear Materials, Vol. 233-237 https://doi.org/10.1016/s0022-3115(96)00010-4	journal	October 1996
High temperature deformation behavior of the molybdenum alloy TZM Mrotzek, T.; Martin, U.; Hoffmann, A. Journal of Physics: Conference Series, Vol. 240 https://doi.org/10.1088/1742-6596/240/1/012079	journal	July 2010
Laboratory Investigation of Vapor Shielding for Lithium-Coated Molybdenum in Devex Jung, Soonwook; Andruczyk, Daniel; Ruzic, David N. IEEE Transactions on Plasma Science, Vol. 40, Issue 3 https://doi.org/10.1109/tps.2011.2181980	journal	March 2012

Cited By (5)

Critical Exploration of Liquid Metal Plasma-Facing Components in a Fusion Nuclear Science Facility Kessel, C. E.; Andruczyk, D.; Blanchard, J. P. Fusion Science and Technology, Vol. 75, Issue 8 https://doi.org/10.1080/15361055.2019.1610685	journal	June 2019
Power handling limit of liquid lithium divertor targets Rindt, P.; Morgan, T. W.; Jaworski, M. A. Nuclear Fusion, Vol. 58, Issue 10 https://doi.org/10.1088/1741-4326/aad290	journal	July 2018
Power handling and vapor shielding of pre-filled lithium divertor targets in Magnum-PSI Rindt, P.; Morgan, T. W.; van Eden, G. G. Nuclear Fusion, Vol. 59, Issue 5 https://doi.org/10.1088/1741-4326/ab0560	journal	March 2019
Using 3D-printed tungsten to optimize liquid metal divertor targets for flow and thermal stresses Rindt, P.; Mata González, J.; Hoogerhuis, P. Nuclear Fusion, Vol. 59, Issue 5 https://doi.org/10.1088/1741-4326/ab0a76	journal	March 2019
Vapor shielding of liquid lithium divertor target during steady state and transient events Marenkov, E.; Pshenov, A. Nuclear Fusion, Vol. 60, Issue 2 https://doi.org/10.1088/1741-4326/ab5eb5	journal	January 2020

Similar Records

Fundamental Surface Science of PFCs for Improved Plasma Performance in NSTX-U

Technical Report · Sun Oct 27 00:00:00 EDT 2019 · OSTI ID:1335696

Koel, Bruce E.

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:1335696

Ono, M

Predictive modeling of a lithium vapor box divertor in NSTX-U using SOLPS-ITER

Journal Article · Tue Jun 01 00:00:00 EDT 2021 · Nuclear Materials and Energy · OSTI ID:1335696

Emdee, E. D.; Goldston, R. J.; Lore, J. D.; +1 more

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
42 ENGINEERING
fusion
divertor
lithium
capillary
NSTX-U

Title: Conceptual design of a pre-loaded liquid lithium divertor target for NSTX-U

Citation Formats

References (27)

Cited By (5)

Similar Records

Related Subjects