Deuterium-plasma studies in the STEP linear device (2018–2019) left the microscopic origin of hydrogen-isotope retention unresolved in three advanced tungsten-based plasma-facing materials. W–V alloys (5–10 wt% V) retain about 6–7 times more deuterium than pure tungsten, but the responsible V-enriched trapping sites and their concentration dependence were never identified. Forged W–Y2O3 shows a texture-dependent desorption peak near 800 K whose assignment to grain boundaries versus Y2O3 particles or oxide/matrix interfaces was never settled. The low retention of W–ZrC was explained only by an unvalidated hypothesis of rapid hydrogen diffusion along grain boundaries. The problem is to identify and quantify the operative trapping sites and transport paths governing deuterium retention in these materials, and to replace these speculative mechanisms with a validated quantitative account (trap inventory plus transport kinetics) consistent with thermal-desorption and retention data across exposure conditions — the basis for tritium-inventory assessment of tungsten divertor materials. Later literature (2021–2025) has substantially clarified the W–ZrC case (ZrC carbon-vacancy trapping, suppression of deuterium-induced cavity propagation by grain-boundary ZrC particles), but the W–V trapping question, the W–Y2O3 interface-trapping question, the W–ZrC transport question, and a predictive retention model remain open.

Background

Tungsten (W) is the leading plasma-facing material (PFM) candidate for the divertors of next-step magnetic-fusion devices because of its high melting point, low sputtering yield, and low hydrogen solubility. A central safety and fuel-cycle constraint on any PFM is hydrogen-isotope retention: trapped tritium accumulates in the material, and the tritium inventory must stay below regulatory limits. Retention is governed by microscopic traps — vacancies, voids, dislocations, grain boundaries (GBs), solute atoms, and second-phase particles — whose binding energies and populations are specific to each material grade.

Because pure W is brittle at operating temperatures, alloyed and dispersion-strengthened W grades were developed, and their deuterium (D) behavior was tested in the STEP linear plasma device ("simulator for tokamak edge plasma"), which reproduces the low-energy, high-flux D plasmas of a divertor that ion-beam studies cannot. A 2019 summary of the STEP program (Yin et al. 2019) reported three system-specific anomalies, each established by thermal desorption spectroscopy (TDS, in which the exposed sample is heated at a controlled rate and the desorbed D is recorded versus temperature, the peak positions encoding trap binding energies) and electron microscopy:

(1) W–V alloys (5 and 10 wt% V; Wang et al. 2018): D-plasma exposure strongly suppressed or fully suppressed surface blistering (dome-shaped surface deformation driven by D accumulation at subsurface defects), but total D retention increased by factors of about 6.2 and 6.9 relative to rolled pure W, attributed only qualitatively to the high hydrogen solubility of vanadium.

(2) High-energy-rate forged (HERF, a high-strain-rate forging route that produces elongated grains and a strong crystallographic texture) W–0.5 wt% Y2O3 (Tan et al. 2018): blistering depended strongly on grain texture — the forging plane (Plane A, equiaxed grains) blistered, while the cross-sectional plane (Plane B, elongated grains with GBs normal to the exposed surface) did not — and TDS of Plane B showed an elevated desorption peak near 800 K, attributed to GB traps but with trapping at Y2O3 particles or oxide/matrix interfaces not experimentally excluded.

(3) W–0.5 wt% ZrC (Liu et al. 2018): D-plasma exposure at a surface temperature of about 453 K produced blisters of order 100 nm (versus micrometre-scale blisters in pure W) together with unexpectedly low total retention; the low retention was explained only by the speculative hypothesis of rapid hydrogen diffusion along GBs or phase interfaces, never directly measured, and the long-term stability of the ZrC particles under combined plasma and neutron irradiation was unknown.

Prior theory makes solute trapping plausible: first-principles calculations (Kong et al. 2016) show that most transition-metal solutes in W, including V, attract hydrogen, with both chemical and elastic contributions. What was missing in 2019 — and remains the open target — is direct identification and quantification of the operative trapping sites and transport paths in each of the three strengthened systems, i.e. the mechanistic information needed to predict tritium inventory rather than only measure it after the fact.

Problem Statement

For the three advanced tungsten-based plasma-facing material systems studied in the STEP linear device — W-V alloys (5-10 wt% V), high-energy-rate forged W-0.5 wt% Y2O3, and W-0.5 wt% ZrC — determine, by direct experiment and/or atomistic simulation validated against experiment, the microscopic deuterium trapping sites and transport mechanisms that govern retention kinetics. Concretely: (i) for W-V, identify the trapping sites responsible for the roughly 6-7-fold retention increase over pure W, determine whether they are V-enriched, and quantify how V concentration and its spatial distribution (homogeneous solute versus segregated particles) control retention and desorption kinetics; (ii) for W-Y2O3, determine whether Y2O3 particles and/or oxide-matrix interfaces act as deuterium traps, and assign the ~800 K Plane-B desorption peak to a specific, experimentally discriminated trap population; (iii) for W-ZrC, establish the mechanism controlling retention — validate or refute rapid grain-boundary/interface diffusion as the cause of the anomalously low retention reported at about 453 K — and provide a quantitative retention-kinetics account under the original STEP exposure conditions. These three sub-questions jointly serve one objective: a validated, quantitative trap-inventory-and-transport account of hydrogen-isotope retention for solute- and dispersion-engineered tungsten, of the kind required for predictive tritium-inventory estimates in fusion-reactor plasma-facing components. The completion boundary is a mechanistic account for each material system that identifies the operative trap sites and transport paths and is consistent with the measured thermal-desorption and retention data across exposure conditions (temperature, fluence).

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