Bulletin of the American Physical Society
63rd Annual Meeting of the APS Division of Plasma Physics
Volume 66, Number 13
Monday–Friday, November 8–12, 2021; Pittsburgh, PA
Session PM10: Mini-Conference: Interactions of Plasmas with Tungsten Surfaces: Experimental StudiesOn Demand
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Chair: Brian Wirth, University of Tennessee Room: Room 406 |
Wednesday, November 10, 2021 2:00PM - 2:30PM |
PM10.00001: Tritium migration and trapping in neutron damaged materials Masashi Shimada, Chase Taylor, Robert D Kolasinski, Jonathan Coburn, Yasuhisa Oya, Yuji Hatano Deuterium-tritium fusion reactions produce 14.1 MeV neutrons that create radiation damages and transmutation elements in fusion reactor materials. A significant amount of solid transmutation elements such as rhenium and osmium will be produced in tungsten (W), one of the candidate plasma-facing component (PFC) materials. A series of thermal-neutron shielded and unshielded neutron-irradiation campaigns was carried out with polycrystalline W and W alloys in High Flux Isotope Reactor (HFIR) under the US-Japan PHENIX and FRONITER projects to advance the understanding of irradiation response of tungsten on thermo-mechanical properties and tritium behavior. Not only the abovementioned solid transmutation elements but also material surfaces and grain structures play critical roles in tritium behavior in neutron-irradiated tungsten as it determines a boundary condition for diffusing tritium. Defect characterization by positron annihilation lifetime spectroscopy and coincidence Doppler broadening measurements, surface compositions, and depth profiles of elements by scanning Auger microscope are being studied prior to prior plasma exposure experiments in thermal-neutron shielded and unshielded neutron-irradiated tungsten. We report defect characterization, surface compositions and depth profiles of elements from unshielded polycrystalline W irradiated at 1073K and 1373K as well as from shielded polycrystalline W irradiated at 873K, 1073K and 1373K. |
Wednesday, November 10, 2021 2:30PM - 3:00PM |
PM10.00002: An Overview of Dispersion-Strengthened Tungsten Alloys as Fusion Plasma-Facing Materials Jean Paul Allain, Xiang Wang, Ashrakat Saefan, Trevor Farris Marchhart, Chase C Hargrove Tungsten (W) is a common plasma-facing component (PFC) material in the divertor region of tokamak fusion devices due to its high melting point and high sputter threshold [1,2]. However, W is intrinsically brittle and is further embrittled under neutron irradiation, and the low recrystallization temperature pose complications in fusion environments [1,3]. In the past decade, the Radiation Surface Science and Engineering Lab (RSSEL) has been developing dispersion-strengthened tungsten composite alloys as an alternative PFC W-based material. These materials have been processed via spark plasma sintering (SPS) with TiC, ZrC, and TaC dispersoids alloyed from 0.5 to 10 weight %. SPS is a powder compaction technique that provides high pressure and heating rates via electrical current, allowing for a lower final temperature and hold time for compaction [4]. Initial testing of material and PMI properties, microstructure, and composition of specimens will be presented. The DS-W materials have shown promise as PFCs indicated by enhanced recrystallization properties at temperatures between 1200-1800 C, oxygen trapping and D and He exhaust management. However, there remains some outstanding challenges and these will be discussed. High-flux exposures at the Magnum-PSI facility at DIFFER exposed samples to a H fluence of 1026 cm-2 and He fluence of 1025 cm-2. In-situ chemistry changes via XPS, ex-situ morphology changes via SEM, and microstructure with TEM will be presented. |
Wednesday, November 10, 2021 3:00PM - 3:30PM |
PM10.00003: Understanding the poloidal distribution of W sourcing by light impurities and heating method in WEST E.A. Unterberg, C.Christopher Klepper, Guido Ciraolo, Corinne Desgranges, Davis Easley, Nicolas Fedorczak, Alberto Gallo, Alex GROSJEAN, Jamie P Gunn, Curtis A Johnson, Yannick Marandet, Emmanuelle Tsitrone Recent experiments were performed in lower-single-null discharges with dominant RF heating up to 5 MW to evaluate in-situ tungsten (W) erosion measurements in an effort to determine the main W sourcing locations within the vessel. The distribution of W sources is seen to change with the increased RF heating, going from dominate gross erosion at either strikepoint (SP) location to a more balanced source distribution between each SP. The main chamber limiter gross erosion is seen to increase with the SOL electron temperature but is ~10x less than the divertor gross erosion, although the penetration probability across the separatrix is higher from this location than from the divertor target locations and potentially has a larger impact on core contamination. Neutral W spectroscopy using W-I emission measurements are taken from multiple poloidal locations to characterize the gross erosion rate around the vessel, and incident particle fluxes are characterized by Langmuir probes, both embedded and plunged, and spectroscopy methods. Constrained by these measurements, SolEdge-EIRENE shows, taking oxygen as a proxy light impurity, consistent poloidal erosion patterns to those measured. These simulations suggest low fractions (~few % of the background electron density) of oxygen are the main erosion drivers throughout the vessel and that the charge state distribution of oxygen has significant impact on the erosion rate. |
Wednesday, November 10, 2021 3:30PM - 3:50PM |
PM10.00004: Characterizing the effects of hydrogen and helium plasmas on tungsten surfaces with in-situ ellipsometry and low-energy ion beam analysis Robert D Kolasinski, Josh A Whaley, Chun-Shang Wong This study includes the use of in-situ techniques to characterize the structure and composition of plasma-exposed surfaces. For analysis of surface structure, we use spectroscopic ellipsometry to quantify the rate of tungsten nanostructure growth during high-flux He plasma exposure. Under appropriate conditions, we have also used ellipsometry to measure erosion of thin oxide layers. For surface composition measurements, we have also developed a high-fidelity in-situ surface technique based on ion scattering spectroscopy that is capable of operating at mTorr pressures. In this case, the surface is probed using a pulsed beam of 1 - 20 keV ions (typically Li+ or Na+ ), and energies of the scattered/recoiled particles are measured using differentially pumped time-of-flight (TOF) detectors. The atomic mass of each surface species can be related to the particle energies using classical kinematics. Both approaches provide insight into the mechanisms governing dynamic surface modification by plasmas. |
Wednesday, November 10, 2021 3:50PM - 4:10PM |
PM10.00005: Experimental determination of spatially-dependent sub-surface impurity content in tungsten exposed to helium plasma operation in WEST and a controlled laboratory plasma source Wendy A Garcia, Elodie Bernard, David Donovan, Jamie P Gunn, Xunxiang Hu, Thierry Loarer, Chad Parish, Jean-Yves Pascal, Bernard Pegourie, Emmanuelle Tsitrone, Ezekial A Unterberg, Brian D Wirth Tungsten (W) will be used for the divertor in ITER and is a leading candidate in future fusion reactors. The divertor will face an extremely hostile environment leading to severe material degradation of W in the form of surface erosion, migration, redeposition, and impurity transport. This imposes a significant challenge on divertor performance and thus, understanding W behaviour under fusion-relevant environment and developing strong diagnostics tools is essential for the development of advanced plasma facing components for future fusion reactors. In this presentation we investigate the depth-dependent helium (He) and oxygen (O) concentration in single crystal W exposed to approximately 1 second of He plasma in the WEST C4 campaign. He exposures range from ~4x1021 to 3.6x1023 m-2. Additionally, polycrystalline W was exposed to a controlled, low flux plasma source for 1 to 10 hours to fluences ranging from 5E19 to 5E21 m-2 at The University of Tennessee at Knoxville(UTK). Comparison between these samples will provide insight to the impact of different plasma conditions, including impurities and plasma flux, on the depth dependent He concentration. He and O have been spatially depth profiled in plasma-exposed W using Laser Ablation Mass Spectrometry (LAMS) in conjunction with Laser Induced Breakdown Spectroscopy (LIBS). The uncertainties of these results will be discussed through detailed introduction of the calibration processes of these two techniques. |
Wednesday, November 10, 2021 4:10PM - 4:30PM |
PM10.00006: Flexural testing and modelling of tungsten by three-point bending for neutron and helium irradiation studies Nathan C Reid, Trevor F Marchhart, Lauren M Garrison, Jean Paul Allain Fusion reactor divertor materials will be exposed to large thermal gradients caused by high heat fluxes. Tungsten materials in the divertor region must be able to withstand the induced thermal stresses with neutron and plasma particle interactions. The mechanical properties need to be investigated with small-scale disk specimen testing due to the low material-volume constrains in neutron irradiation testing environments. This study seeks to determine the stresses in tungsten under bending when it has been exposed to fission-based neutron irradiation and surface ion damage. The maximum stress in bending is located at the surface of the specimen, allowing to examine the surface under peak stresses. Modelling was performed to determine how these stresses accumulate on an unroughened surface compared to a surface that has an altered morphology from surface ion damage. Polycrystal and single crystal tungsten with orientation in the (110) plane at the disk surface were irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory and tested with three-point bend. The irradiations were performed in the removable beryllium layer of HFIR with a gadolinium thermal neutron shield to doses of 0.2-0.7 DPA in temperature zones of 430-670, 740-960, and 880-1090 C. |
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