Quantitatively measuring the influence of helium in plasma-exposed tungsten

AbstractTungsten samples are exposed to 3He plasma to quantify their helium retention behavior. The retention saturates quickly with helium fluence and increases only slightly from 4.3×1019He/m2 at 773K, to 7.5×1019He/m2 at 973K. The helium content increases dramatically to 6.8×1020He/m2 when fuzz is formed on the surface of a sample exposed at 1173K, but the majority of the retained helium (5.1×1020He/m2) is found to reside below the layer of fuzz tendrils. Additional tungsten samples were exposed to either simultaneous, or sequential, D/He plasma, followed by TDS. Measurements show the majority of the D retained during simultaneous exposures is located in the near surface region of helium nano-bubbles. No deuterium was detected in any of the samples after the heating to 1273K, but 67% of the helium was released from simultaneously exposed samples, and only 23% of the helium was released from the sequentially exposed samples

Implications of PMI and wall material choice on fusion reactor tritium self-sufficiency

Tritium self-sufficiency is a critical issue for the production of nuclear fusion energy. Here we quantify the impact of co-deposition of eroded wall material and fuel on the tritium particle balance in a hypothetical reactor system. The expected ITER plasma parameters and geometry are used to estimate the amount of eroded material from a full tungsten, beryllium or carbon device. Measured D concentrations in co-deposits are extrapolated to the wall temperature expected in future reactors and used along with these eroded flux estimates to determine the net loss probability of tritium from the device due to co-deposition. The use of liquid divertor surfaces is also considered with the amount of tritium residing in the recirculating liquid estimated. The general conclusion, from a tritium self-sufficiency viewpoint, is that one should avoid low-Z materials that readily form hydrogen bonds, in favor of high-Z non-hydride forming materials. Keywords: Plasma-material interactions, Co-deposition, Tritium, Self-sufficienc

Deuterium retention in re-solidified tungsten and beryllium

Leading edges of the ITER tungsten (W) divertor are expected to melt due to transient heat loads from edge localized modes (ELMs), and melting of the entire divertor surface will occur during vertical displacement events (VDEs) and disruptions. In addition, understanding tritium retention in plasma facing materials is critical for the successful operation of ITER due to safety reasons. Thus, the question of how melting affects hydrogenic retention is highly relevant for fusion devices. Here we use an Nd:YAG laser to melt tungsten and beryllium in vacuo, and the samples are subsequently exposed to deuterium plasma with sample temperatures ranging from 370 to 750 K. The deuterium content in re-solidified and reference (no laser) samples is measured using thermal desorption spectroscopy and modeled using TMAP-7. In all cases, the re-solidified samples have lower retention compared to the reference samples. For re-solidified tungsten, the most significant effect is in the 1.8 eV trap with peak thermal desorption temperature of ∼750 K, which had a 77% reduction in the peak release rate compared with the reference sample. SEM imaging indicates that laser melting and re-solidification of tungsten anneals intrinsic defects that act as nucleation sites for larger-scale defects that develop during plasma exposure. However, melting does not significantly affect traps with lower de-trapping energies of 1.0 eV and 1.4 eV. In beryllium, melting and cracking results in lower retention compared to the reference sample by 40%, and thermal desorption profiles indicate that the diffusion depth of deuterium into re-solidified beryllium is lower than that of the reference sample. Keywords: Tokamak plasma-material interactions, Transient heating, Laser melting, Hydrogen retentio

A model of ballistic helium transport during helium-induced fuzz growth in tungsten

Ballistic helium transport in tungsten during fuzz growth was investigated to provide insight into the helium-induced nanostructuring process which adversely affects the near-surface properties of tungsten plasma-facing components. An analytical model of helium ion transport in the vacuum region within the fuzz layer was developed, whereby He ions are assumed to move in straight-line trajectories with mean free paths determined from fuzz/nanotendril dimensions. Reflection of ions from tendril sides was considered, with He ions allowed to implant into the bulk at the base of the fuzz layer if they maintain ≥5eV of energy necessary to overcome the He–W surface barrier potential, resulting in an increase in the effective He mean free path. Using the model results for helium implantation in the bulk, a helium fluence-fuzz thickness relation was achieved, which matches well with experimental data in the literature, and implies that the fuzz growth rate is consistent with ballistic implantation into the bulk through the growing porous fuzz layer