One of the main advantages of using tungsten (W) as a plasma facing material (PFM) is its low uptake and retention of tritium. However, in high purity (ITER grade) W, hydrogenic retention increases significantly with neutron-induced displacement damage in the W lattice. This experiment examines an alternative W grade PFM, ultra-fine grain (UFG) W, to compare its retention properties with ITER grade W after 12 MeV Si ion displacement damage up to 0.6 dpa (displacements per atom.) Following exposure to plasma in the DIII-D divertor, D retention was then assessed with Nuclear Reaction Analysis (NRA) depth profiling up to 3.5 µm and thermal desorption spectrometry (TDS). Undamaged specimens were also included in our test matrix for comparison. For all samples, D release peaks were observed during TDS at approximately 200 °C and 750 °C. For the ITER-grade W specimens, the intensity of the 750 °C release peak was more pronounced for specimens that had been pre-damaged.
We present analysis and modeling of Al sputtering and ionization in attached, low-power L-mode plasmas near the outer divertor strike point of the DIII-D tokamak. Al serves as a useful proxy for Be, the low-Z main wall material for ITER and JET that will undergo significant divertor plasma contact upon migrating from the first wall to the divertor. Al is easily distinguishable from background sources in DIII-D (namely C and B), has a high physical sputtering yield similar to Be, and has a long ionization mean free path compared to its gyro radius ({λi} /rgyro ∼ 2.5). Using neutral Al emission imaging techniques, we measured a toroidal and radial asymmetry in the shape of the photo-emission plumes of sputtered neutral Al that was consistent with previously observed asymmetry in the distribution of redeposited Al in these experiments. We propose that the main cause of the emission and redeposition asymmetry is due to a sputtering anisotropy caused by near-grazing angle incident ions. The observed emission asymmetry was reproduced using a simple emission/ionization model that included full angular distributions of sputtering yield and energy calculated by SDTRIM.SP, but not when symmetric, mono-energetic cosine sputtering distributions were assumed. We used an ion orbit tracking model to calculate the distributions of ion impact energies through the potential gradient in the magnetic pre-sheath and Debye sheath. We found that with the magnetic field pitch angle (1.5°-2° with respect to the surface plane), the majority of ions strike the surface at <15° with respect to the surface plane, leading to angular sputtering yield and energy distributions with significant forward-scattering bias. We also observed surface microstructure consistent with directional sputtering and ion flux shadowing expected from the calculated ion incidence angles.
Additively manufactured (AM) austenitic stainless steels are intriguing candidates for the storage of gaseous hydrogen isotopes because complex vessel geometries can be built more easily than by using conventional machining options. Parts built with AM stainless steel tend to have excellent mechanical properties (with tensile strength, ductility, fatigue crack growth, and fracture toughness comparable to or exceeding that of wrought austenitic stainless steel). However, the solidification microstructures produced by AM processing differ substantially from the microstructures of wrought material. Some features may affect permeability, including some amount of porosity and a greater amount of ferrite. Because the diffusivity of hydrogen in ferrite is greater than in austenite (six orders of magnitude at ambient temperature), care must be taken to retain the performance that is taken for granted due to the base alloy chemistry. Furthermore, AM parts tend to have greater dislocation densities and greater amounts of carbon, nitrogen, and oxygen. These features, along with the austenite/ferrite interfaces, may contribute to greater hydrogen trapping. We report the results of our studies of deuterium transport in various austenitic (304L, 316, and 316L) steels produced by AM. Manufacturing by Powder Bed Fusion (PBF) and two different blown powder methods are considered here (Laser Engineered Net Shaping® (LENS®) and a Direct Laser Powder Deposition (DLPD) method with a higher laser power)). The hydrogen permeability (an equilibrium property) changes negligibly (less than a factor of 2), regardless of chemistry and processing method, when tested between 150 and 500°C. This is despite increases in ferrite content up to FN=2.7. However, AM materials exhibit greater hydrogen isotope trapping, as measured by permeation transients, thermal desorption spectra, and inert gas fusion measurement. The trapping energies are likely modest (<10 kJ/mol), but may indicate a larger population of trap sites than in conventional 300-series stainless steels.
In this work, we examine the response of an ultra-fine grained (UFG) tungsten material to high-flux deuterium plasma exposure. UFG tungsten has received considerable interest as a possible plasma-facing material in magnetic confinement fusion devices, in large part because of its improved resistance to neutron damage. However, optimization of the material in this manner may lead to trade-offs in other properties. We address two aspects of the problem in this work: (a) how high-flux plasmas modify the structure of the exposed surface, and (b) how hydrogen isotopes become trapped within the material. The specific UFG tungsten considered here contains 100 nm-width Ti dispersoids (1 wt%) that limit the growth of the W grains to a median size of 960 nm. Metal impurities (Fe, Cr) as well as O were identified within the dispersoids; these species were absent from the W matrix. To simulate relevant particle bombardment conditions, we exposed specimens of the W-Ti material to low energy (100 eV), high-flux (> 1022 m− 2 s− 1) deuterium plasmas in the PISCES-A facility at the University of California, San Diego. To explore different temperature-dependent trapping mechanisms, we considered a range of exposure temperatures between 200 °C and 500 °C. For comparison, we also exposed reference specimens of conventional powder metallurgy warm-rolled and ITER-grade tungsten at 300 °C. Post-mortem focused ion beam profiling and atomic force microscopy of the UFG tungsten revealed no evidence of near-surface bubbles containing high pressure D2 gas, a common surface degradation mechanism associated with plasma exposure. Thermal desorption spectrometry indicated moderately higher trapping of D in the material compared with the reference specimens, though still within the spread of values for different tungsten grades found in the literature database. For the criteria considered here, these results do not indicate any significant obstacles to the potential use of UFG tungsten as a plasma-facing material, although further experimental work is needed to assess material response to transient events and high plasma fluence.
To address the transport and trapping of hydrogen isotopes, several permeation experiments are being pursued at both Sandia National Laboratories (deuterium gas-driven permeation) and Idaho National Laboratories (tritium gas- and plasma-driven tritium permeation). These experiments are in part a collaboration between the US and Japan to study the performance of tungsten at divertor relevant temperatures (PHENIX). Here we report on the development of a high temperature (≤1150 °C) gas-driven permeation cell and initial measurements of deuterium permeation in several types of tungsten: high purity tungsten foil, ITER-grade tungsten (grains oriented through the membrane), and dispersoid-strengthened ultra-fine grain (UFG) tungsten being developed in the US. Experiments were performed at 500–1000 °C and 0.1–1.0 atm D2 pressure. Permeation through ITER-grade tungsten was similar to earlier W experiments by Frauenfelder (1968–69) and Zaharakov (1973). Data from the UFG alloy indicates marginally higher permeability (< 10×) at lower temperatures, but the permeability converges to that of the ITER tungsten at 1000 °C. The permeation cell uses only ceramic and graphite materials in the hot zone to reduce the possibility for oxidation of the sample membrane. Sealing pressure is applied externally, thereby allowing for elevation of the temperature for brittle membranes above the ductile-to-brittle transition temperature.
A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.