Session UM09: Mini-Conference: Characterization of Modeling of Plasma-Facing Materials for ITER and Beyond

Beryllium Surface Reconstructions and Intermixing on Tungsten

Tungsten and Beryllium are plasma facing materials (PFMs) chosen as the diverter and first wall material, respectively, at ITER’s tokamak fusion reactor. The Be walls have been shown to erode under the intense conditions of the reactor and Be can deposit on the W diverter. A significant buildup of Be can lead to intermixing and the formation of Be-W alloys. These alloys have significantly lower melting temperatures and different thermal conductivity compared to pure tungsten which could lead to reduced diverter lifetimes and impact the safety of the reactor. The first step in understanding Be-W alloy formation is the assessment of Be intermixing into W. 

Density functional theory is used to discover optimal Be structures as a function of coverage on W(110) and W(211) surfaces. On W(110), a densely packed hexagonal surface reconstruction, which constitutes a full single layer, has been found. Below this density, a superposition of this high and a low-density (1x1) coverage is seen. The W(211) surface has deep troughs which fill with Be and, at the highest single layer coverage, Be atoms are stacked in twisted columns close to the [011] direction. The properties and relative stability of these structures are analyzed using a work function calculation, surface energies and comparison to experimental images. Intermixing is shown to be energetically unfavorable with less than or equal to a full single layer coverage which qualitatively agrees with experimental findings that a critical amount of Be build up is required for significant intermixing and alloying   

Strain and thermal gradient effects on the transport properties of intrinsic defects and impurities in tungsten

Plasma-facing materials (PFMs) in a fusion reactor are expected to withstand stringent conditions, with high heat and particle fluxes that create strong gradients of temperature and concentration of diverse species. These species will then migrate in the presence of the afore-mentioned gradients and large strain fields. In this work, we use nonequilibrium molecular dynamics (NEMD) simulations to study the transport properties of H, He, and SIAs in the presence of a thermal gradient and different strain fields in tungsten. The NEMD simulations reveal that defects and impurity atoms tend to migrate toward the hot regions of the material (negative heat of transport). The resulting concentration profiles are in agreement with the predictions of irreversible thermodynamics. Furthermore, strain seems to play a critical role in the transport of these species significantly changing the concentration profiles. We demonstrate that the resulting steady-state profiles significantly depend on these fields.   

Investigations on dynamic wall recycling from tungsten divertor material in DIII-D

Several tungsten samples were simultaneously exposed in the DIII-D lower divertor to ELMy low power H-mode attached plasmas using the Divertor Material Evaluation System (DiMES). Atomic and molecular deuterium recycling fluxes from tungsten samples were monitored through the measurements of spatially and ELM-resolved atomic Dα and molecular Fulcher-band emission [1]. The net variations of the molecular and atomic deuterium recycling fluxes observed above the various tungsten samples and the surrounding graphite head are similar during ELMs. Those results indicate that the nature of the wall material has virtually no effect on dynamic fuel recycling in the low power plasma conditions encountered in those experiments. In particular, two tungsten samples were pre-exposed to helium particles with an incidence ion energy about 60 eV and a fluence about  resulting in the formation of helium nano-bubble layers, suggesting that micro-structural defects weakly impact dynamic wall recycling in the considered plasma conditions. Modeling of deuterium recycling during an ELM from pristine and pre-loaded tungsten materials is performed with the cluster dynamics code Xolotl and assessed against experimental observations. Finally, we briefly discuss the observed toroidal shift of the molecular deuterium emission clouds with respect to the location of the tungsten samples, which may provide insights into the toroidal transport of deuterium molecules. 

 

[1] Bykov, I., et al. Physica Scripta 2020.T171 (2020)

    

Effect of Helium Flux on Helium Accumulation in the Near-surface Region of Plasma-Exposed Tungsten

Tungsten is a candidate material for plasma-facing components (PFCs) in nuclear fusion reactors. During its operation, PFCs are subjected to temperatures > 800 K and a low-energy ( eV), high-flux (on the order of ) of He+ ion irradiation, resulting in subsurface He accumulation. To understand the subsurface accumulation of implanted helium as a function of its flux, object kinetic Monte Carlo (OKMC) simulations were carried 100 eV — helium implantation of tungsten (100) surface were performed. These simulations were performed at a te2mperature of 933 K for He implantation fluxes ranging from  to . In the near-surface region, helium clusters elastically interact with the free surface. The interaction is attractive and becomes increasingly stronger, thereby lowering the activation energy barriers for helium clusters to hop towards the surface and trap mutate with decreasing depth. Furthermore, the lowering of the trap mutation barrier with decreasing depth also results in the kick-out of multiple tungsten atoms, which sometimes is followed by the partial dissociation of the mutated helium cluster. Therefore, these simulations were performed using not only the depth-dependent helium cluster migration and trap mutation barriers but also the depth-dependent trap mutation processes. The types of trap mutation processes and their energetics were obtained using molecular dynamics simulations. A detailed discussion on the influence of helium flux on subsurface helium and, helium bubble accumulation as a function of depth and fluence will be presented.   

Response of Dispersion-Strengthened Tungsten Alloys to Helium Irradiation

Helium ion irradiation is a major threat to the integrity of tungsten materials as plasma facing components, since the generated He bubbles can substantially degrade the material mechanical, thermal, and surface properties. In this talk, we will focus on the response of dispersion-strengthened tungsten (DS-W) alloys to helium irradiation. DS-W is a new class of W alloys strengthened by carbide dispersoids, such as TiC, ZrC and TaC. Compared to pure W, DS-W possesses increased recrystallization temperature and improved ductility, making DS-W promising candidates as divertor materials of future fusion reactors. Our studies showed that He bubble formation in DS-W was suppressed by the carbide dispersoids. In DS-W alloys irradiated at 800°C to the fluence of 1×10²⁴ He/m², no bubbles were founded within the carbide particles or at the carbide-W interface, while obvious bubbles formed in the W matrix. Similar phenomenon was observed using in-situ He irradiation conducted at 950°C, suggesting the capability of carbide dispersoids to trap He atoms as small clusters. Thermal desorption spectroscopy analysis discovered that the thermal release of implanted He decreased as carbide concentration increased from zero to 10 wt.% in DS-W. Microscale mechanical testing found that TiC-strengthened W showed better mechanical performance than pure W and other DS-W alloys after He irradiation. Our study provides a new avenue to designing W alloys with superior tolerance to He irradiation by optimizing the dispersoid concentration and distribution.        

Overview of the new IGNIS-2 surface science facility for the in-situ analysis of fusion plasma-facing components at Penn State

The IGNIS-2 in-situ surface science facility is a new experimental platform that will be utilized for the analysis of plasma-facing components (PFC). This overview will discuss the design and commissioning of IGNIS-2, and ongoing projects which utilize the facility's unique capabilities to study reconstituted materials and liquid metal-surface interactions for PFC development. IGNIS-2 will utilize broad-beam ion sources to modify the surface of materials while simultaneously measuring the evolution of the surface properties. This new facility will continue to feature Low Energy Ion Scattering Spectroscopy and X-ray Photoelectron Spectroscopy (XPS) to measure chemical composition at the surface and in the near-surface region, while a major upgrade to a monochromatic X-ray source will greatly improve the resolution of XPS measurements. The new facility will also incorporate optical techniques, such as Multi-beam Optical Stress Sensor and Ellipsometry, to make in-situ measurements of surface stress, film thickness, and surface roughness. The main analysis chamber will be connected to other chambers through a series of in-vacuo components, allowing samples to be moved to chambers dedicated other purposes such as thin-film deposition via evaporation or magnetron sputtering. A major research activity on IGNIS-2 is the study of liquid lithium films on nanostructured porous tungsten substrates, supporting work on the NSTX-U spherical tokamak. Another major focus of research on IGNIS-2 is the study of reconstituted boron/tungsten layers, as would be found in a fusion reactor with tungsten PFCs and boron wall-conditioning due to continuous erosion and redeposition of material at the surface of PFCs.   

Investigating bubble formation in dispersion-strengthened tungsten alloys using in-situ TEM irradiation

The divertor region of tokamak reactors experience an extremely harsh environment, including the intense irradiation flux, the very high heat flux and the high temperature. The leading material of choice is tungsten because of its very high melting point. However, pure tungsten itself accumulates damage quickly from particle irradiation, so new tungsten alloys need to be developed to withstand such an environment. Dispersion-strengthened tungsten (DS-W) is a class of tungsten alloys strengthened by transition metal carbides such as TiC, ZrC, and TaC. Initial studies have found that DS-W alloys show better long-term use in highly damaging environments as they possess increased recrystallization temperature and improved ductility. Helium (He) bubble formation is a serious threat to materials within the divertor region. Here we investigated bubble formation in DS-W using in-situ He irradiation in a transmission electron microscope (TEM). The W-TaC sample was irradiated at 950°C with a He flux of 3*10¹² cm ^-2 s ^-1. Our analysis showed He bubble formation was suppressed by the carbides within the DS-W alloys. During the irradiation, the bubble area number density began at 226 /µm² at 994 s into irradiation and gradually increased to a maximum density of 689 /µm² at 2117 s, and then falling to 544 /µm² at 2329 s into the irradiation. This in-situ TEM videos showed He bubbles preferred to nucleate and grow along the grain boundaries in the W matrix, and bubbles did not appear to form within the TaC dispersoids or at the dispersoid-matrix interfaces.   

Recent work in dispersion-strengthened tungsten alloys as plasma-facing component materials in future fusion reactors

Tungsten is the current material of choice for plasma-facing component envisioned in future nuclear fusion reactor systems. The divertor region of future nuclear fusion reactors will be exposed to extremely plasma fluence of 10²⁶ m^-2 and temperatures in excess of 2000°C. These conditions require materials, such as tungsten and its alloys and composites, with exceptional thermo-mechanical properties. A high melting point, low coefficient of thermal expansion, high sputtering resistance, and high conductivity make tungsten an ideal candidate. In addition, PMI properties such as He exhaust management and hydrogen retention must be addressed.  Despite its desirable traits, sustained plasma exposure can lead to significant microstructural damage in tungsten. In addition, tungsten alloys are susceptible to a high ductile-to-brittle transition temperature and recrystallization under irradiation. To overcome these limitations, the Radiation Surface Science and Engineering Lab (RSSEL) has been developing dispersion-strengthened tungsten (DS-W) alloys. In DS-W, the tungsten matrix is strengthened by carbide particles such as TiC, ZrC and TaC. Our studies showed that DS-W alloys possess improved structure stability and increased radiation resistance compared to the pure tungsten counterpart. Continuing investigations into plasma-material interactions (PMIs) of DS-W and the resulting surface, microstructural and mechanical changes will be presented.