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DE-SC0021005: Advanced Ultrafine Dispersion-Strengthened Tungsten Alloys as Burning Plasma-Facing Component Materials

Award Status: Active
  • Institution: The Pennsylvania State University, University Park, PA
  • DUNS: 003403953
  • Most Recent Award Date: 10/20/2023
  • Number of Support Periods: 4
  • PM: Echols, John
  • Current Budget Period: 09/01/2023 - 08/31/2024
  • Current Project Period: 09/01/2023 - 08/31/2025
  • PI: Wang, Xing
  • Supplement Budget Period: N/A

Public Abstract

Renewal: Advanced ultrafine dispersion-strengthened tungsten alloys as burning plasma-facing
component materials
Jean Paul Allain, Pennsylvania State University (Principal Investigator)
Xing Wang, Pennsylvania State University (Co-Investigator)
Project Summary
Tungsten (W) is the plasma-facing component (PFC) material of choice for ITER in the divertor region
because of its favorable thermomechanical properties, such as a high melting point, low sputtering threshold,
and excellent thermal conductivity. However, W also suffers from serious limitations as PFCs, including a
high ductile-to-brittle transition temperature, radiation-induced embrittlement, and recrystallization at
normal tokamak operating temperatures. In the previous program, we focused on the discovery and
development of ultrafine dispersion-strengthened tungsten (UDS-W) alloys, which are self-healing and
adaptive materials for the PMI (plasma-material interface) envisioned for future plasma-burning extreme
environments in thermonuclear fusion reactors. Fabricated using the advanced spark plasma sintering
technique, UDS-W has shown inhibited recrystallization at 1800 °C, enhanced ductility at lower
temperatures, and suppressed helium bubble formation at the W-dispersoid interfaces. Our results indicate
that UDS-W provides a promising strategy to significantly improve the mechanical properties, radiation
tolerance, and PMI performance of W.
UDS-W is an excellent example showing how advanced manufacturing and complex material component
design could bring transformative changes to plasma-facing materials. However, when planning for future
tokamak reactors like DEMO and a Fusion Pilot Plant (FPP), it is realized that the synergistic impact from
the high thermal load, neutron damage, and plasma exposure creates unrivalled challenges for materials.
As pointed out in the 2021 National Academy of Engineering (NAE) report Bring Fusion to the U.S. Grid,
research is required to “move beyond the early stage of alloy and composite development to examining the
material performance and degradation in the complex neutron, plasma material, and thermal-mechanical
loading conditions.”
This proposal directly addresses the request from the NAE report by focusing on systematically
investigating the separate and synergistic effects of neutron displacement damage, high thermal
loading conditions, and helium/deuterium plasma exposure on two key PMI properties of UDS-W,
i.e., retention of hydrogen isotopes and resistance to surface nanostructuring. We propose studies that
proceed from relatively simple single variable experiments to more complex, fully integrated, multiplevariable
tests. This proposal also supports the DOE FES Burning Plasma Science: Long Pulse—Materials
mission by three primary objectives: (1) Elucidate the fundamental role of W-dispersoid interfaces on
hydrogen isotopes and helium retention; (2) Establish an understanding of microstructure evolution under
displacement damage and its impact on PMI properties; (3) Evaluate surface stability of UDS-W under high
heat loading and reactor relevant conditions.
The work in this proposal combines fundamental and discovery research that uncover process-compositionproperty-
function relations of novel tungsten-based PFC materials. The proposal provides a unique
perspective at the intersection of materials science and PMI. The proposed PMI studies aim to fill
significant PMI knowledge gaps identified in the DOE Department of Energy Fusion Energy Sciences
Workshop on Plasma Materials Interactions 2015. These gaps were articulated in research themes that
emphasized the need to gain insight into the reconstituted surfaces in PMI under reactor-relevant long-pulse
prototypical environments. This gap is addressed in this proposal by carefully examining the incubation
regimes of irradiation-driven defects and correlating to effects on PMI properties and their evolution over
very large fluence (i.e., prototypical long-pulse) regimes. In particular, this proposal investigates the
characteristic behavior of surface composition and morphology of UDS-W materials developed with spark
plasma sintering and the properties intrinsic in UDS-W materials to survive these harsh environments.

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