Session JM11: Mini-Conference: Measuring and Modeling of Plasma Material Interactions I

Laser Blow Off and Impurity Entrainment in CSDX

Material migration in tokamaks is especially important when considering the lifetime of plasma facing components. However, the flow geometry in the scrape off layer can be complex and diagnostic access is limited. To study the impurity transport in a simple geometry, a laser blow off apparatus was installed on the Controlled Shear Decorelation eXperiment (a 3m long linear helicon source operated plasma machine with an electron temperature of ~4 eV and density of ~10$^{13}$ per cm$^3$). The parallel velocity and diffusion coefficients for the impurity are determined by modeling the impurity injection with the 1-D advection-diffusion equation. The parallel velocity of the impurity cloud found with the model agrees very well with laser induced fluorescence (LIF) measurements of the parallel plasma ion velocity. This indicates that the bismuth impurities are fully entrained in the plasma, which is consistent with classical collisional theories.   

Characterization of hydrogen binding to tungsten and beryllium surfaces using low energy ion beam analysis

In this study, we use low energy ion beam analysis to determine how hydrogen interacts with tungsten and beryllium surfaces. The goal of this work is to provide insight into processes that contribute to recycling from plasma-facing surfaces in magnetic fusion devices. Here we have applied low energy ion scattering (LEIS) to enable detection of adsorbed hydrogen at sub-monolayer resolution and to provide isotopic sensitivity. We probe the surfaces of interest with He$+$ and Ne$+$ at energies less than 5 keV to determine the structure and composition of the first few atomic layers. This approach enables us to examine how hydrogen surface concentrations evolve in real time, providing insight into adsorption kinetics. In addition, we have developed a means of determining the hydrogen binding configuration at different temperatures by exploiting mechanisms of ion channeling along surfaces. Using these methods, we have been able to identify hydrogen binding configurations for the W(100)$+$H, W(110)$+$H, and Be(0001)$+$H adsorption systems. We also report on our efforts to more accurately and efficiently model atomic collisions during scattering, key steps needed to extract structural information from LEIS signals.   

Tritium Retention and Permeation in Ion- and Neutron-Irradiated Tungsten under US-Japan PHENIX Collaboration

A critical challenge for long-term operation of ITER and beyond to a FNSF, a DEMO and future fusion reactor will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to intense heat and neutral/ion particle fluxes under the extreme fusion nuclear environment, while minimizing in-vessel inventories and ex-vessel permeation of tritium. Recent work at Tritium Plasma Experiment demonstrated that tritium diffuses in bulk tungsten at elevated temperatures, and can be trapped in radiation-induced trap site (up to 1 at. {\%} T/W) in tungsten [M. Shimada, et.al., Nucl. Fusion 55 (2015) 013008]. US-Japan PHENIX collaboration (2013--2019) investigates irradiation response on tritium behavior in tungsten, and performs one-of-a-kind neutron-irradiation with Gd thermal neutron shield at High Flux Isotope Reactor, ORNL. This presentation describes the challenge in elucidating tritium behavior in neutron-irradiated PFCs, the PHENIX plans for neutron-irradiation and post irradiation examination, and the recent findings on tritium retention and permeation in 14MeV neutron-irradiated and Fe ion irradiated tungsten.   

Effect of Damaging Temperature on Deuterium Retention in Tungsten

Fusion-relevant displacement damage in W and its influence on D retention is explored. Ferroni identified three recovery stages for damaged W near 623, 913, and 1253 K [1], and post-damage annealing at elevated temperature before or during plasma exposure has shown a reduction in D retention [2]. In this work 2 and 5 MeV Cu ions were first used to produce up to 0.2 dpa damage in W samples under various temperatures ranging from 773 to 1273 K that were then exposed to D plasma at 383 K to a fluence of 10$^{24}$ ions/m$^{2}$. Subsequent Nuclear Reaction Analysis and Thermal Desorption Spectrometry show that increased temperature during damage creation reduces D retention more than published post-damage annealing [2]. Experimental results and initial modeling work will be reported. \\[4pt] [1] F. Ferroni et al., Acta Materialia 90, 380-393 (2015)\\[0pt] [2] M.H.J. 't Hoen et al., Nucl. Fusion 53, 043003 (2013)   

Retention modeling in ion damaged W and diffusivity calculation including trapping effects

A Cu ion beam is used to induce controlled levels of damage (10$^{-3}$, 10$^{-2}$, and 10$^{-1}$ dpa) in room temperature W samples. A single 5 MeV beam energy yielding a peaked damage profile 0.8 $\mu $m into the material, or 3 beam energies (0.5, 2, and 5 MeV) producing a relatively uniform damage profile from the near surface up to 1 $\mu $m were used. The W samples were then exposed to a D plasma ion fluence of 10$^{24}$ ions/m$^{2}$ at 383 K, and the resulting D retention was measured using the D($^{3}$He,p)$\alpha $ reaction analysis. We observe that there is no significant difference in retention whether the damage profile is peaked or uniform, further justifying the use of heavy ions as neutron surrogates. A retention model [1] provides concentration profiles that can be directly compared to NRA data and total retention measurements. Taking the trapping energies from DFT calculations, the only free-parameter is the defect density. The model can fit our data within the experimental error of the measurements. A new diffusion coefficient is calculated with the model that is not only a function of temperature but also a function of the trapped concentration. This calculation resolves discrepancies of various diffusivity measurements and models in the literature.\\[4pt] [1] J.L. Barton et al., J. Nucl. Mater., 463 (2015) 1129-1133   

Revisiting reaction-diffusion model of thermal desorption spectroscopy experiments on hydrogen retention in material for fusion applications

Plasma-material interactions may strongly influence plasma performance and life-time of future magnetic fusion devices. Understanding the multifaceted physics of hydrogen retention in plasma-facing components (PFC) is thus crucial, but remains challenging due to the wide spectrum of retention processes on PFC surface and in PFC bulk induced by long-time exposure of PFC to high flux of energy and particles [1]. We revisit here some aspects of reaction-diffusion models used to investigate hydrogen retention in material. We focus on analysis of thermal desorption spectroscopy (TDS) experiment considering only one type of traps in material and first neglecting surface effects. We show that solute hydrogen concentration in retention region usually remains in equilibrium during TDS experiments. In this regime, analytic description of thermal desorption spectra indicates that trapping of solute hydrogen during TDS cannot be ignored. Main features of thermal desorption are then analytically described and refined interpretation of Arrhenius plots is proposed. Effects of surface processes on hydrogen outgassing during TDS experiments are then introduced and surface-limited outgassing regimes are discussed. \\[4pt] [1] J. Roth and K. Schmid, Phys. Scr., 014031 (2011)   

Object Kinetic Monte Carlo Simulations of Radiation Damage In Bulk Tungsten

Results are presented for the evolution of radiation damage in bulk tungsten investigated using the object KMC simulation tool, KSOME, as a function of dose, dose rate and primary knock-on atom (PKA) energies in the range of 10 to 100 keV, at temperatures of 300, 1025 and 2050 K. At 300 K, the number density of vacancies changes minimally with dose rate while the number density of vacancy clusters slightly decreases with dose rate indicating that larger clusters are formed at higher dose rates. Although the average vacancy cluster size increases slightly, the vast majority exists as mono-vacancies. At 1025 K void lattice formation was observed at all dose rates for cascades below 60 keV and at lower dose rates for higher PKA energies. After the appearance of initial features of the void lattice, vacancy cluster density increased minimally while the average vacancy cluster size increases rapidly with dose. At 2050 K, no accumulation of defects was observed over a broad range of dose rates for all PKA energies studied in this work. Further comparisons of results of irradiation simulations at various dose rates and PKA spectra, representative of the High Flux Isotope Reactor and future fusion relevant irradiation facilities will be discussed.