Lauren M. Garrison

Flux threshold measurements of He-ion beam induced nanofuzz formation on hot tungsten surfaces

We report measurements of the energy dependence of flux thresholds and incubation fluences for He-ion induced nano-fuzz formation on hot tungsten surfaces at UHV conditions over a wide energy range using real-time sample imaging of tungsten target emissivity change to monitor the spatial extent of nano-fuzz growth, corroborated by ex situ SEM and FIB/SEM analysis, in conjunction with accurate ion-flux profile measurements. The measurements were carried out at the multicharged ion research facility (MIRF) at energies from 218 eV to 8.5 keV, using a high-flux deceleration module and beam flux monitor for optimizing the decel optics on the low energy MIRF beamline. The measurements suggest that nano-fuzz formation proceeds only if a critical rate of change of trapped He density in the W target is exceeded. To understand the energy dependence of the observed flux thresholds, the energy dependence of three contributing factors: ion reflection, ion range and target damage creation, were determined using the SRIM simulation code. The observed energy dependence can be well reproduced by the combined energy dependences of these three factors. The incubation fluences deduced from first visual appearance of surface emissivity change were (2–4) × 1023 m−2 at 218 eV, and roughly a factor of 10 less at the higher energies, which were all at or above the displacement energy threshold. The role of trapping at C impurity sites is discussed.

Progress in the U.S./Japan PHENIX Project for the Technological Assessment of Plasma Facing Components for DEMO Reactors

Abstract The PHENIX Project is a 6-year U.S./Japan bilateral, multi-institutional collaboration program for the Technological Assessment of Plasma Facing Components for DEMO Reactors. The goal is to address the technical feasibility of helium-cooled divertor concepts using tungsten as the armor material in fusion power reactors. The project specifically attempts to (1) improve heat transfer modeling for helium-cooled divertor systems through experiments including steady-state and pulsed high-heat-load testing, (2) understand the thermomechanical properties of tungsten metals and alloys under divertor-relevant neutron irradiation conditions, and (3) determine the behavior of tritium in tungsten materials through high-flux plasma exposure experiments. The High Flux Isotope Reactor and the Plasma Arc Lamp facility at Oak Ridge National Laboratory, the Tritium Plasma Experiment facility at Idaho National Laboratory, and the helium loop at Georgia Institute of Technology are utilized for evaluation of the response to high heat loads and tritium interactions of irradiated and unirradiated materials and components. This paper provides an overview of the progress achieved during the first 3 years and discusses the plan for the remainder of the project.

New Insight into Gridded Inertial Electrostatic Confinement (IEC) Fusion Devices

Abstract Gridded inertial electrostatic confinement (IEC) devices use a 10-200 kV voltage difference to accelerate ions through a 0.1-10 mTorr background gas in a spherical or cylindrical geometry. The detailed investigation of a gridded IEC device using DD fuel has resulted in several surprises that have greatly altered our perception of how these systems operate. It was found that there are at least 4 major misconceptions that have been in place for over 15 years on how such IEC systems operate. These misconceptions range all the way from what energetic ion is causing the majority of fusions, to the energy and charge state of the reacting ions. Experimental results will illustrate some of the surprising reactions that are taking place in DD gridded system.

Surface Pore Formation in Helium Implanted Fine-Grain Tungsten and Tungsten Needles as Engineered First Wall and Divertor Plate Materials

Abstract Surface morphology changes of sub-micron tipped tungsten needles (W.N.) and an engineered fine-grain tungsten (FGW) were studied after implantation with He ions at reactor relevant conditions. Surface and subsurface pore formation was observed on all of the samples by using scanning electron microscopy (SEM) and focused ion beam (FIB) milling. Additionally, helium retention analysis was performed on the FGW and compared to several previously studied W materials. Three samples of FGW were irradiated with 30 keV 3He ions to 3×1017 He+/cm2 at 700 °C, 9×1017 He+/cm2 at 850 °C, and 1×1019 He+/cm2 at 1050 °C. SEM analysis revealed that the threshold for visible pore formation was below ˜1018 He+/cm2. The sample irradiated to the highest fluence showed “coral-like” morphology on the surface, and FIB analysis showed that the sub-surface semi-porous layer extended almost one micron below the surface. The percentage of implanted helium retained in the samples ranged from 4.5-40%. Two W.N. were implanted with 100 keV 4He ions to conditions of 3×1018 He+/cm2 at 700 °C and 1.3x1019 He+/cm2 at 1000 °C. Extensive pore formation was observed on both specimens. FIB analysis revealed that a sub-surface semi-porous layer developed after ion implantation that extended ˜300 nm in the W.N. implanted to the lower dose, and over 1500 nm in the needle implanted to the higher dose. This second needle also exhibited a drastic morphology change, which appears to be a result of the unraveling of the grains at the tip.

Recent Advances in IEC Physics and Technology at the University of Wisconsin

Abstract The University of Wisconsin-Madison has conducted research on gridded inertial electrostatic confinement (IEC) devices for the past 18 years. There are currently 4 experimental devices operating at voltages up to 180 kV and 60 mA. These devices have uncovered several new phenomena that have greatly improved our understanding of IEC devices. Recent advances include the discovery of a significant negative ion component of DD plasmas and spatial profiles of fusion reactions that did not conform to our prior understanding of these devices. The use of this technology has also contributed to our understanding of surface damage to high temperature in-vessel W components after even low exposures to energetic He ion fluences. Expansion of the voltage-ion current parameter space to 300 kV-200 mA in the near future will help our understanding of advanced fusion fuel cycles.

The Influence of Microstructure on Deuterium Retention in Polycrystalline Tungsten

Abstract The retention of hydrogen isotopes in the plasma-facing materials of a fusion reactor is dependent on the density of trapping sites in the material. One factor that can influence the trapping defects is the surface state of the material before exposure. Mechanically polished, electropolished, and recrystallized tungsten samples were compared by exposing them to 350 eV D+ beams with peak fluences of ~1 × 1024 D+/m2 at 500 and 740 K at the Multicharged Ion Research Facility (MIRF). At the exposure temperature of 740 K, no significant retention was detected. For material exposed at 500 K, significant differences in retention were observed, and the order of increasing retention was recrystallized, electropolished, and mechanically polished. The other variable besides surface treatment was the time delay between ion exposure and thermal desorption spectroscopy which also may have impacted the retention measurements if there was out-gassing of the D while samples were in storage before thermal desorption spectroscopy (TDS).