Chase N. Taylor

Development of positron annihilation spectroscopy for characterizing neutron irradiated tungsten

Tungsten samples (6 mm diameter and 0.2 mm thick) were irradiated to 0.025 and 0.3 dpa with neutrons in the High Flux Isotope Reactor at Oak Ridge National Laboratory as part of the US/Japan Tritium, Irradiation and Thermofluids for America and Nippon (TITAN) collaboration. Samples were then exposed to deuterium plasma in Idaho National Laboratorys Tritium Plasma Experiment at 100, 200 and 500 °C to a total fluence of 1 × 1026 m−2. Nuclear reaction analysis (NRA) and Doppler broadening positron annihilation spectroscopy (DB-PAS) were performed at various stages to characterize radiation damage and retention. We present the first results of neutron irradiated tungsten characterized by DB-PAS in order to study defect concentration. Two positron sources, 22Na and 68Ge, probe ~58 μm and through the entire 200 μm thick samples, respectively. DB-PAS results reveal clear differences between the various irradiated samples. These results, and a correlation between DB-PAS and NRA data, are presented.

Tritium Plasma Experiment Upgrade and Improvement of Surface Diagnostic Capabilities at STAR Facility for Enhancing Tritium and Nuclear PMI Sciences

Abstract Recently, Tritium Plasma Experiment (TPE), a unique high-flux linear plasma device that can handle beryllium, tritium and neutron-irradiated plasma facing materials, has undergone major upgrades in its electrical and control systems. The upgrade has improved worker occupational safety, and enhanced TPE plasma performance to better simulate extreme plasma-material-interaction (PMI) conditions expected in ITER, Fusion Nuclear Science Facility (FNSF) and demonstration fusion power plant (DEMO). The PMI determines a boundary condition for diffusing tritium into bulk plasma-facing components (PFCs) and plays critical role in in-vessel and ex-vessel safety assessments. Enhancing surface capabilities for tritium-contaminated and radioactive samples is crucial for the PMI sciences in burning plasma long pulse operation. The TPE Upgrade and improvement of surface diagnostic capabilities for tritium-contaminated and radioactive samples at STAR facility help enhance tritium and nuclear PMI sciences for the development of reliable PFCs and tritium fuel cycle in ITER, FNSF and DEMO.

Softening due to Grain Boundary Cavity Formation and its Competition with Hardening in Helium Implanted Nanocrystalline Tungsten

The unique ability of grain boundaries to act as effective sinks for radiation damage plays a significant role in nanocrystalline materials due to their large interfacial area per unit volume. Leveraging this mechanism in the design of tungsten as a plasma-facing material provides a potential pathway for enhancing its radiation tolerance under fusion-relevant conditions. In this study, we explore the impact of defect microstructures on the mechanical behavior of helium ion implanted nanocrystalline tungsten through nanoindentation. Softening was apparent across all implantation temperatures and attributed to bubble/cavity loaded grain boundaries suppressing the activation barrier for the onset of plasticity via grain boundary mediated dislocation nucleation. An increase in fluence placed cavity induced grain boundary softening in competition with hardening from intragranular defect loop damage, thus signaling a new transition in the mechanical behavior of helium implanted nanocrystalline tungsten.

Direct depth distribution measurement of deuterium in bulk tungsten exposed to high-flux plasma

Understanding tritium retention and permeation in plasma-facing components is critical for fusion safety and fuel cycle control. Glow discharge optical emission spectroscopy (GD-OES) is shown to be an effective tool to reveal the depth profile of deuterium in tungsten. Results confirm the detection of deuterium. A ∼46 μm depth profile revealed that the deuterium content decreased precipitously in the first 7 μm, and detectable amounts were observed to depths in excess of 20 μm. The large probing depth of GD-OES (up to 100s of μm) enables studies not previously accessible to the more conventional techniques for investigating deuterium retention. Of particular applicability is the use of GD-OES to measure the depth profile for experiments where high deuterium concentration in the bulk material is expected: deuterium retention in neutron irradiated materials, and ultra-high deuterium fluences in burning plasma environment.

Deuterium Retention in Helium and Neutron Irradiated Molybdenum

Abstract Understanding and managing D retention in plasma facing components is essential for tritium safety in fusion reactors. Neutron irradiated and virgin low carbon arc cast (LCAC) Mo, as well as Mo foil samples with and without He pre-irradiation, were used to investigate D retention. D and He retention were investigated simultaneously in thermal desorption spectroscopy using a high resolution residual gas analyzer. Results show a significant increase in D retention with He pre-irradiation. Vacancies and vacancy clusters are found to retain D in LCAC samples, but neutron irradiated Mo retains more D in vacancy clusters.