Dynamic interactions between energetic D and He ions on lithium-tungsten plasma-facing interfaces

Neff, Anton L.

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https://hdl.handle.net/2142/99298 Copy

Description

Title

Dynamic interactions between energetic D and He ions on lithium-tungsten plasma-facing interfaces

Author(s)

Neff, Anton L.

Issue Date

2017-10-06

Director of Research (if dissertation) or Advisor (if thesis)

Allain, Jean Paul

Doctoral Committee Chair(s)

Allain, Jean Paul

Committee Member(s)

• Ruzic, David
• Stubbins, James
• Bellon, Pascal
• Hattar, Khalid

Department of Study

Nuclear, Plasma, & Rad Engr

Discipline

Nuclear, Plasma, Radiolgc Engr

Degree Granting Institution

University of Illinois at Urbana-Champaign

Degree Name

Ph.D.

Degree Level

Dissertation

Keyword(s)

• Plasma Facing Component
• Lithium
• Tungsten
• Thin Film
• Plasma Surface Interactions

Abstract

One of the best options for providing clean abundant energy for the needs of the world is nuclear power. Nuclear fission is a well established method to produce energy without any greenhouse gases but nuclear fusion has the potential to produce even more energy per volume of fuel compared with nuclear fission and will produce less radioactive by-products. One promising configuration is thermonuclear magnetically confined fusion. As improvements continue in understanding and enabling confinement of highly-dense high-temperature fusion plasmas, a considerably higher demand is placed on the plasma facing components (PFCs) with an increase of fusion power production. Currently within operating experimental reactors, the PFCs need to handle only a few MW/m$^2$ of heat flux and particle fluxes of $\sim$10$^{22}$ ions/m$^2$s but with the construction of the International Thermonuclear Test Reactor (ITER), materials will need to tolerate two order of magnitude higher particle fluxes and heat fluxes over 10 MW/m$^2$. This is primarily due to the expected production of a positive net power output because of the higher plasma temperature and density achieved with the larger size of ITER. These conditions will only worsen for PFCs as fusion research progresses toward a high duty-cycle, reactor-relevant, burning plasma fusion energy source. The material that is slated for use in the divertor of ITER, where the particle and heat loads will be the highest, is tungsten (W). Tungsten was selected because it has good thermal properties and a high sputter threshold. In spite of these beneficial properties, the surface morphology of tungsten, when exposed to deuterium (D) and helium (He), will change to form blisters, bubbles, nano-structured ``fuzz'', etc. that reduces the thermal conductivity and changes the sputtering properties. These changes, if they cause erosion of W, can contaminate the plasma, which may lead to a plasma disruption. One material that has shown the ability to improve plasma performance, and has a lower chance of disrupting the plasma due to erosion, is lithium (Li). When Li has been used as a PFC, the plasma performance has improved because the lithium has reduced the recycling of fuel from the walls. It is also beneficial to have Li in a tokamak (fusion plasma chamber) because it is low atomic number (Z) and the plasma can tolerate a high percentage of Li impurities. Additionally, an isotope of Li, $^6$Li can produce tritium (T) when it interacts with neutrons, where T is a radioactive, low natural abundance isotope of H used as fuel. This research investigates how the combination of Li and W will behave as a tokamak PFC. Specifically, we investigated, with simultaneous and sequential D and He irradiations at both low and high flux conditions, whether He ions reduce D retention in Li films on W by breaking chemical complexes between Li, D, and oxygen (O) or by preventing the formation of these complexes. With the high flux irradiations when He was introduced, the He has been observed to reduce the D retention in the films due to the presence of WO$_2$ peaks, under the passivated lithium layer. These tungsten (II) oxides point to fewer Li-O-D complexes allowing Li to reduce the oxide state of the W. With the low flux sequential irradiations, \textit{in-operando} X-ray Photoelectron Spectroscopy (XPS) scans of the O 1s region showed an increase in the ratio between Li-O-D and Li$_2$O with D following He and a reduction in the same ratio with He following D. These results show that the He breaks a chemical complex between Li, O and D. The use of He to reduce the fuel retention has the potential to create a method to control fuel recycling from the walls. Another effect that was tested was the interaction between Li and the nano-structured fuzz that tungsten will develop when it is exposed to low energy He ions at high fluxes ($>$10$^{22}$ ions/m$^2$s) and at surface temperature above 900$^\circ$C. The tests have shown that the fuzz layer thickness does not change when a Li layer is present before plasma exposure, so it does not prevent the fuzz from forming. However, secondary ion mass spectrometry (SIMS) depth profiles show that Li is not completely removed and that the secondary ion counts of Li are maintained, dropping by less than an order of magnitude when the fuzz is present and drops by one order of magnitude in half the depth when there are only He bubbles formed on the surface. This reduction indicates that the fuzz prevents the escape of Li during irradiation. Additionally, this result led to investigating the erosion behavior of fuzz with Li coatings. The erosion tests have shown an increase in tendril thicknesses with a 2X increase without Li and a 4X increase with Li but little change in the layer thickness. Erosion measurements indicate little reduction in the erosion of W fuzz with Li coatings among the fuzz. The XPS signal from the W 4f peaks showed a 13\% drop in W eroded when Li is deposited on fuzz and only a 3\% drop when Li is deposited before fuzz is formed. The persistence of the Li among the fuzz tendrils encourages the study of microscopically porous materials as reservoirs for low Z metals and liquid metals as prospective PFCs.

Graduation Semester

2017-12

Type of Resource

text

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Copyright 2017 Anton L Neff

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