Outer Divertor Damage Characterization from Deuterium Plasma Bombardment in Graphene-Coated Tungsten

FUSION SCIENCE AND TECHNOLOGY · VOLUME 75 · 542–550 · AUGUST 2019 © 2019 American Nuclear Society DOI: https://doi.org/10.1080/15361055.2019.1610317
Outer Divertor Damage Characterization from Deuterium Plasma Bombardment in Graphene-Coated Tungsten in the C-2W Device
Marcos X. Navarro, *a Marziyeh Zamiri,b Martin E. Griswold,c John F. Santarius,a Gerald L. Kulcinski,a Max Lagally,a and Toshiki Tajimac
aUniversity of Wisconsin-Madison, Madison, Wisconsin bUniversity of Ottawa, Ottawa, Ontario, Canada
cTAE Technologies, Lake Forest, California
Received June 14, 2018
Accepted for Publication April 11, 2019
Abstract — This research explores the performance of graphene as a coating for plasma-facing components (PFCs) in a nuclear fusion environment. Our recent studies have shown that graphene can act as a resistant layer against plasma exposure and ion bombardment. PFCs tend to develop surface morphologies that lead to mass loss of the wall material, potentially diminishing their lifetime and degrading plasma performance. We present a characterization of graphene-coated samples of W irradiated in the C-2W divertor. Energy analyzers were used to determine average ion fluxes to the
18 + 2
samples on the order of 10 D /cm . Two samples were exposed over 1210 plasma discharges. Raman
spectroscopy showed that slow ions (30 < E < 100 eV) interact strongly with the graphene, introducing vacancies into the membrane (ID/IG ~ 0.7), making it possible to assess the limiting factors on such a coating’s lifetime. We also found that graphene slows down impurity deposition on the material surfaces due to graphene’s stable configuration and low surface energy. This first attempt at testing the coating in a large-scale fusion experiment aims to expand the possible wall candidates for PFCs.
Keywords — Graphene, tungsten, deuterium, divertor, C-2W.
Note — Some figures may be in color only in the electronic version.
I. INTRODUCTION
An excellent balance between potential reactor
attractiveness and technical development risk motivates
the development of field-reversed configuration (FRC)
1,2
Technologies) FRC experimental facility,3–5 known both as C-2W and Norman, is shown in Fig. 1. The FRC is an approximately ellipsoidal magnetic configuration immersed inside the magnetic field lines of an open- ended magnetic geometry. The open field lines guide charged particles toward the inner or end divertors for carrying heat and removing impurities from the system. Therefore, it is important to characterize the damage to the divertor surfaces due to the particle and heat fluxes.
The need for material candidates for plasma-facing
components (PFCs) is rising because of their potential to
limit plasma performance. In the presence of plasmas
containing helium, PFCs tend to develop surface
morphologies that lead to mass loss of the divertor wall
material, potentially diminishing its lifetime and affecting
fusion power plants.
geometry facilitates the design of shields, magnets, input- power systems, and tritium-breeding blankets, while the high FRC β (≡plasma pressure/magnetic field pressure) increases the plasma power density and allows a compact fusion core. The surface heat flux to the divertor is moderate despite a high-power density because the plasma flowing to the end chamber walls carries much of the charged-particle fusion power. The computer-aided design (CAD) of the Tri Alpha Energy (TAE
The linear, cylindrical FRC
6,7
properties, tungsten has become a leading candidate for
the safety of a fusion reactor.
Because of its refractory
*E-mail: navarrogonza@wisc.edu 542