About

Dr. David J. Sprouster is an Assistant Professor in the Department of Materials Science and Chemical Engineering at Stony Brook University. His research focuses on characterizing and controlling matter over multiple length- and time-scales. Professor Sprouster is actively developing and engineering next generation energy materials that can withstand the extreme environments common to advanced fission and fusion reactor technologies. His group employs multi-modal synchrotron- and neutron-based characterization methods to support applied and fundamental research programs. He received his Ph.D. in Physics from The Australian National University in 2010. Prior to joining Stony Brook University as a faculty member, he was a Senior Scientist and Assistant Research Professor at Stony Brook University with a joint appointment in the Nuclear Reactor Laboratory at the Massachusetts Institute of Technology.

Research topics

• Metallurgy
• Materials science
• Composite material
• Physics
• Crystallography
• Optics
• Nanotechnology
• Chemistry

Selected publications

• High Temperature Stabilization of Ultrafine Grain Tungsten Alloys Through Synergistic Compositional Complexities

  Acta Materialia · 2026-05-01

  article
  PublisherDOI

• Understanding The Impact of Laser Powder Bed Fusion Processing on the Sensitization of Austenitic 316L Stainless Steel Using Standardized Testing Procedures

  Research Square · 2026-01-14

  preprintOpen access
  PublisherOA PDFDOI

• Low temperature neutron irradiation stability of Zirconium hydride and Yttrium hydride

  Journal of Nuclear Materials · 2025-03-18 · 8 citations

  articleOpen access1st authorCorresponding

  Metal hydrides, including ZrH x and YH x , are of particular interest for advanced thermal fission reactors as they have high neutron moderating power and can be used at relatively high temperatures. They have direct applications as core components including as a moderating addition in nuclear fuel, and as neutron reflectors or moderators. Understanding their thermal and irradiation-induced property changes are important to their engineering application. Specifically, evolving metal hydrogen ratios are of critical importance. In this work we discuss the post-irradiation examination of neutron irradiated ZrH 2-x and YH 2-x specimens. We employ multiple characterization techniques including X-ray diffraction, scanning electron microscopy and thermophysical (thermal diffusivity) to determine the irradiation-induced macro- and microstructural evolution as a function of irradiation temperature. We readily quantify degradations in the thermal diffusivity, changes in lattice parameters, and an increase in metallic Zr indicative of hydrogen release in ZrH 2-x specimens. Interestingly, minimal-to-nil change in the metallic Y fraction was quantifiable in the YH 2-x specimens and modest changes in the thermal diffusivity occur for the temperature and dose studied. The loss of hydrogen in the ZrH 2-x samples is related to an apparent irradiation-accelerated desorption of hydrogen by the high ionizing radiation components (gamma, epithermal and fast neutron fluxes) from the in-core neutron irradiation. The most apparent feature from the microstructural analysis for both metal hydrides was a temperature-dependent decrease in the X-ray diffraction peak broadening, attributable to changes in the number and makeup of the two-dimensional defects. These results and trends improve both the fundamental understanding of neutron-solid interactions, and the development of such an important class of core materials.

  PublisherDOI

• Effect of stress on Laves phase precipitation in creep ruptured Grade 92 ferritic martensitic steel characterized by a novel accessible method

  Acta Materialia · 2025-09-18 · 2 citations

  article
  PublisherDOI

• An Investigation of Phase Evolution in Isothermally Aged Alloys 690, 625, and 625 plus

  2025-08-10

  article

  Abstract The formation of long-range ordered phases has long been debated as a potential concern in nuclear power applications. The Ni2Cr phase is known to cause hardening, embrittlement and matrix lattice contraction in Ni-Cr alloys, but the timescale associated with its formation in commercial alloys is unclear. In this study, we isothermally age alloys 690, 625, and 625 plus (625+) at four temperatures, 330, 360, 418, and 475 °C, for up to 50,000 hours. Samples are characterized by synchrotron x-ray diffraction and microhardness. Alloy 690 shows matrix lattice contraction, short range ordering, and a small amount of hardening (~6 %) at 418 and 475 °C, but no long-range ordered peaks after 50k hours due to its high Fe content. Thus, Ni2Cr long-range ordering is sluggish in alloy 690 at LWR-relevant temperatures and further timepoints are needed to quantify its potential formation and effects on mechanical properties. The synchrotron x-ray diffraction patterns do not reveal the long-range ordered Ni2Cr phase in alloy 625 after 50k hrs but significant matrix lattice contraction and noticeable hardening (~28 %) occur at the 475 °C temperature, which saturate after 10k hrs. The synchrotron x-ray diffraction patterns also do not show the long-range ordered Ni2Cr phase in alloy 625+ after 50k hrs, but the matrix and γ” phases gradually contract and harden (~15 %) after 50k hrs at 475 °C, thought to relate to γ’ formation.

  PublisherDOI

• Microstructure and Mechanical Behavior of a Tic Nanoprecipitate Strengthened V Alloy

  SSRN Electronic Journal · 2025-01-01

  preprintOpen access
  PublisherDOI

• Bonding of vanadium- and Iron-based alloys as interlayers for plasma-facing and structural materials in fusion systems

  Materials & Design · 2025-09-22

  articleOpen access

  • L2 1 intermetallic phase found at FeCrAl–VCrAl interface. • Interface hardening directly linked to L2 1 phase. • Additively manufactured builds showed B2 phase, unlike sintered samples. • Cr buffer layer proposed to avoid intermetallic formation. • No σ-phase detected in any tested sample across the interface. Vanadium alloys and FeCrAl were investigated as interlayers between tungsten and reduced activation ferritic martensitic steel for fusion system components to avoid formation of intermetallic phase at operating temperatures between 550 and 1100 °C, while maintaining a body centered cubic phase throughout the interface. Physical and mechanical properties need to be graded between tungsten and steel, but recent results showed a significant hardness increase at the FeCrAl to vanadium alloy interface. Here, a sintered sample of these alloys was annealed for extended time, and the microstructure was investigated to provide a better understanding of the phenomena. A comparison with an additively manufactured interface of the same material is provided. An unexpected L2 1 intermetallic phase formation has been revealed using microscopy and synchrotron techniques and will inform future additive manufacturing approaches of the interface. A Cr layer interface as a preliminary solution was proposed between the Vanadium alloy and FeCrAl alloy interface.

  PublisherDOI

• Historic and modern nuclear graphite impurities: Pathways to improved waste strategies

  Current Opinion in Solid State and Materials Science · 2025-12-01 · 2 citations

  articleOpen accessSenior author

  • Survey of impurity content, including 14 C-producting nitrogen, in graphite from original CP-1 AGOT through currently used nuclear grades. • Sources and distribution of nitrogen in artificial graphite identified. • Recommendation made for obtainable nitrogen levels made to reduce cost of irradiated graphite waste disposal. Graphite has been used in large volumes as a structural material and neutron moderator since the earliest days of nuclear fission. However, no international consensus exists on the disposal of irradiated graphite, leaving much of the historic radioactive graphite inventory in interim vault or silo storage. With several new graphite-moderated reactors planned or under construction, the issue of graphite waste management is becoming increasingly urgent. This paper reviews and quantifies impurities in both historic and modern nuclear graphite, with emphasis on nitrogen—responsible for much of the 14 C inventory—and chlorine, which plays a critical role in repository performance and design. Modern graphites, benefitting from stringent quality-control measures developed for non-nuclear industries, meet or exceed the ASTM Ultra-High Purity nuclear standards, even without halide purification. Both chlorine and nitrogen concentrations have declined over time. For chlorine, identified as a key impurity influencing U.S. waste repository design, we propose a target of 0.1 appm in as-fabricated billets as a reasonable benchmark. Nitrogen sources are traced throughout the graphite production process, with surface and bulk concentrations characterized for all materials studied. Modern graphites commonly exhibit nitrogen levels below 5 appm, with values approaching 1 appm achievable. Using such reduced-nitrogen grades is critical to keeping graphite-induced radioactivity below the greater-than-Class-C waste threshold, thereby avoiding disposal cost penalties of nearly an order of magnitude.

  PublisherDOI

• Interplay of surface energy and rheology in biopolymer soil enhancement

  Polymer Testing · 2025-10-25

  articleOpen access

  Biopolymers such as xanthan gum (XG) and locust bean gum (LBG) hold great potential as eco-friendly alternative soil binders. In this work, we investigated the impact of XG/LBG mixtures on the unconfined compressive strength (UCS) of sand. The high strength of dry biopolymer/sand arises from the cohesion between solid polymer films and sand particles which supported by work of adhesion calculation and soil mechanics measurement. LBG exhibits much lower sand reinforcement efficacy because polymers unevenly distributed within sand matrix. The formation of a core-shell structure in LBG/sand is an interplay of surface free energy and viscoelastic properties of polymer solutions. This structure is altered when LBG mixed with XG at varying ratios as those physical properties changed due to the complexity of polymer chains association. By probing these factors, we aim to elucidate the role of surface energies and polymer physics in governing the strength of the sand/polymer network, thereby contributing to a more comprehensive understanding polymer-sand interface. The low strength of gels (G’ ∼10Pa) cannot solely account for the increased UCS of wet sand over 10 kPa. Instead, the high strength of biopolymer/sand is more likely derived from the granular particles with biopolymers as solid glue. • Addition of as little as 0.2 wt% of Xanthan gum greatly enhances the compressive strength of sand granules. • Additional enhancement is obtained when Xanthan Gum is mixed with Locust Bean Gum at different ratios. • LBG's low surface energy drives it to the air interface, forming core shell structure that weakens the compressive strength. • Electron microscopy and micro-CT show polymer films and fibrils coating the sand granules, explaining the higher compressive modulus.

  PublisherDOI

• High-temperature nanoindentation creep studies on castable and sintered nanostructured low-activation ferritic-martensitic alloys

  Journal of Nuclear Materials · 2025-04-04 · 4 citations

  articleOpen accessSenior authorCorresponding

  • Creep properties of CNA & SNA are similar due to MX and M23C6 precipitates. • Nanoindentation creep tests reveal similar properties due to microstructure. • Stress exponent (n) of 8–33 indicates dislocation-dominated creep in both alloys. • SNA by DCS replicates CNA's microstructure & creep properties for rapid production. • Findings aid high-stress, high-temperature materials for nuclear applications. In this article, we present the creep characteristics of two reduced activation ferritic-martensitic steels of identical starting compositions formed by different fabrication routes: a nanostructured ferritic alloy commonly referred to as a castable nanostructured alloy (CNA) and a sintered nanostructured alloy (SNA) variant. Through a series of nanoindentation experiments spanning a temperature range of 25 °C to 650 °C, with a maximum load of 100 mN, we find creep behaviors in the cast and sintered materials to be remarkably similar. The creep stress exponent ( n ) for CNA and SNA were found to be in the range of 8–35 and the activation volume was ∼14–42 b 3 , underscoring a dominance of dislocation-mediated mechanisms in both alloys. Notably, we observed a decline in the creep stress exponent with increasing temperature, attributable to the heightened influence of thermally activated dislocations. This phenomenon suggests a potential transition in the deformation mechanism towards a thermally activated dislocation climb process, significantly impacting the observed creep behavior.

  PublisherDOI

Frequent coauthors

• Felipe Kremer

  Australian National University

  103 shared

• C. J. Glover

  Pfizer (United States)

  101 shared

• Sahar Mirzaei

  100 shared

• R. Feng

  100 shared

• M. C. Ridgway

  Australian National University

  93 shared

• S. Decoster

  IMEC

  92 shared

• Scott A. Medling

  90 shared

• Salvy P. Russo

  90 shared

Education

• Ph.D.

  The Australian National University

  2010

• Other, Senior Scientist/Assistant Research Professor

  Stony Brook University