Stuart Maloy

Profile Information

Name

Dr. Stuart Maloy

Institution

Pacific Northwest National Laboratory

Position

Senior Nuclear Materials Adviser

Affiliation

ANS, TMS

h-Index

ORCID

0000-0001-8037-1319

Biography

Stuart Maloy is a senior nuclear materials adviser in the Reactor Materials and Mechanical Design Group at Pacific Northwest National Laboratory. He has previous experience at Los Alamos National Laboratory where he worked for 33 years specializing in radiation effects in materials. He is also a National Laboratory Professor with the nuclear engineering department at the University of New Mexico and a fellow of the American Nuclear Society. He has a Bachelors (’89) Masters (’91) and PhD (’94) in Materials Science from Case Western Reserve University and is a registered PE in Metallurgy.

He has applied his expertise to characterizing and testing the properties of metallic and ceramic materials in extreme environments such as under neutron and proton irradiation at reactor relevant temperatures. This includes testing the mechanical properties (fracture toughness and tensile properties) of Mod 9Cr-1Mo, HT-9, 316L, 304L, Inconel 718, Al6061-T6 and Al5052 after high energy proton and neutron irradiations using accelerators and fast reactors. Characterization of materials after testing includes using transmission electron microscopy for analyzing defects such as dislocations, twins and second phases, using high resolution electron microscopy to characterize defects at an atomic level and nanoscale mechanical testing. Stuart has >250 peer reviewed technical publications (>6000 citations, H-index of 42) and numerous presentations.

Expertise

Alloys, Characterization, Mechanical Testing, Steel

Publications:

		"A Novel Approach for Manufacturing Oxide Dispersion Strengthened (ODS) Steel Cladding Tubes using Cold Spray Technology" Ben Maier, Mia Lenling, Stuart Maloy, Kumar Sridharan, Nuclear Engineering and Technology Vol. 2019 Link A novel fabrication method of oxide dispersion strengthened (ODS) steel cladding tubes for advanced fast reactors has been investigated using the cold spray powder-based materials deposition process. Cold spraying has the potential advantage for rapidly fabricating ODS cladding tubes in comparison with the conventional multi-step extrusion process. A gas atomized spherical 14YWT (Fe-14%Cr, 3%W, 0.4%Ti, 0.2%Y, 0.01%O) powder was sprayed on a rotating cylindrical 6061-T6 aluminum mandrel using nitrogen as the propellant gas. The powder lacked the oxygen content needed to precipitate the nanoclusters in ODS steel, therefore this work was intended to serve as a proof-of-concept study to demonstrate that free-standing steel cladding tubes with prototypical ODS composition could be manufactured using the cold spray process. The spray process produced an approximately one-millimeter thick, dense 14YWT deposit on the aluminum-alloy tube. After surface polishing of the 14YWT deposit to obtain desired cladding thickness and surface roughness, the aluminum-alloy mandrel was dissolved in an alkaline medium to leave behind a free-standing ODS tube. The as-fabricated cladding tube was annealed at 1000 °C for 1 hour in an argon atmosphere to improve the overall mechanical properties of the cladding.
		"Atom probe study of irradiation-enhanced a' precipitation in neutron-irradiated Fe–Cr model alloys" Wei-Ying Chen, Jian Gan, Stuart Maloy, Kun Mo, Maria Okuniewski, James Stubbins, Yinbin Miao, Yaqiao Wu, Carolyn Tomchik, Journal of Nuclear Materials Vol. 462 2015 242-249 Link Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation doses and irradiation temperatures on α′ phase formation in Fe–Cr model alloys (10–16 at.%) irradiated at 300 and 450 °C to 0.01, 0.1 and 1 dpa. For 1 dpa specimens, α′ precipitates with an average radius of 1.0–1.3 nm were observed. The precipitate density varied significantly from 1.1 × 1023 to 2.7 × 1024 1/m3, depending on Cr concentrations and irradiation temperatures. The volume fraction of α′ phase in 1 dpa specimens qualitatively agreed with the phase diagram prediction. For 0.01 dpa and 0.1 dpa, frequency distribution analysis detected slight Cr segregation in high-Cr specimens, but not in Fe–10Cr specimens. Proximity histogram analysis showed that the radial Cr concentration was highest at the center of α′ precipitates. For most precipitates, the Cr contents were significantly lower than that predicted by the phase diagram. The Cr concentration at precipitate center increased with increasing precipitate size.
		"Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ˜295 °C to ˜6.5 dpa" Stuart Maloy, G. Robert Odette, Tarik Saleh, Takuya Yamamoto, Osman Anderoglu, T. J. Romero, Shuangchen Li, James Cole, Randall Fielding, Journal of Nuclear Materials Vol. 468 2016 232-239 Link Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσy) and reductions in uniform strain ductility (eu) are observed, where the latter can be understood in terms of the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are also analyzed to establish relations between Δσy and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σyu). The latter shows that higher σyu correlates with lower Δσy. In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher eu than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. Notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.
		"Correlation of in-situ transmission electron microscopy and microchemistry analysis of radiation-induced precipitation and segregation in ion-irradiated advanced ferritic/martensitic steels" Ce Zheng, Stuart Maloy, Djamel Kaoumi, Scripta Materialia Vol. 162 2019 460-464 Link
		"Direct comparison of nanoindentation and tensile test results on reactor-irradiated materials" David Krumwiede, Takuya Yamamoto, Tarik Saleh, Stuart Maloy, G. Robert Odette, Peter Hosemann, Journal of Nuclear Materials Vol. 504 2018 135-143 Link
		"Effect of dose on irradiation-induced loop density and Burgers vector in ion-irradiated ferritic/ martensitic steel HT9" Ce Zheng, Stuart Maloy, Djamel Kaoumi, Philosophical Magazine Vol. 98 2018 2440-2456 Link TEM samples of F/M steel HT9 were irradiated to 20 dpa at 420°C, 440°C and 470°C in a TEM with 1 MeV Kr ions so that the microstructure evolution could be followed in situ and characterized as a function of dose. Dynamic observations of irradiation-induced defect formation and evolution were done at different temperatures. The irradiation-induced loops were characterized in terms of their Burgers vector, size and density as a function of dose and similar observations and trends were found at the three temperatures: (i) both a/2 <111> and a <100> loops are observed; (ii) in the early stage of irradiation, the density of irradiation-induced loops increases with dose (0-4 dpa) and then decreases at higher doses (above 4 dpa), (iii) the dislocation line density shows an inverse trend to the loop density with increasing dose: in the early stages of irradiation the pre-existing dislocation lines are lost by climb to the surfaces while at higher doses (above 4 dpa), the build-up of new dislocation networks is observed along with the loss of the radiation-induced dislocation loops to dislocation networks; (iv) at higher doses, the decrease of number of loops affects more the a/2 <111> loop population; the possible loss mechanisms of the a/2 <111> loops are discussed. Also, the ratio of a <100> to a/2 <111> loops is found to be similar to cases of bulk irradiation of the same alloy using 5 MeV Fe2+ions to similar doses of 20 dpa at similar temperatures.
		"In situ neutron diffraction study on temperature dependent deformation mechanisms of ultrafine grained austenitic Fe-14Cr-16Ni alloy" Bjorn Clausen, Stuart Maloy, Cheng Sun, Kaiyuan Yu, International Journal of Plasticity Vol. 53 2014 125-134 Link
		"Influence of injected interstitials on the void swelling in two structural variants of 304L stainless steel induced by self-ion irradiation at 500 °C" Cheng Sun, Frank Garner, Lin Shao, Xinghang Zhang, Stuart Maloy, Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms Vol. 409 2017 323-327 Link
		"In-situ observation of nano-oxide and defect evolution in 14YWT alloys" Osman El Atwani, Meimei Li, Stuart Maloy, Eda Aydogan, Materials Characterization Vol. 170 2020 110686 Nanostructured ferritic alloys (NFAs) are considered as candidates for structural components in advanced nuclear reactors due to their excellent radiation resistance as a result of a high density of nano-oxides (NOs) in the microstructure. Therefore, gaining an understanding on the stability of NOs under irradiation is crucial. In this study, we have investigated the evolution of defects and NOs in 14YWT NFAs under in-situ Kr ion irradiation at room temperature (RT) and 450 °C up to 10 dpa. It has been found that irradiations at 450 °C do not create any changes in the NOs, similar to the bulk irradiations. On the other hand, elemental mapping indicates that NOs dissolve mostly after 10 dpa irradiations at RT. Thus, while defects are both annihilated and pinned by NOs at low doses (before the dissolution of NOs), glissile loops start to escape to the foil surface at high doses (after the dissolution of NOs), justifying the significantly low fraction of <111> loops compared to the literature values. High resolution transmission electron microscopy analysis has shown that the NOs are mostly coherent Y2Ti2O7 particles with pyrochlore crystal structure after both RT and 450 °C irradiations, similar to those observed before irradiation.
		"In-situ radiation response of additively manufactured modified Inconel 718 alloys" Eda Aydogan, Osman El Atwani, Begum Erdem, Wei-Ying Chen, Meimei Li, ARUN DEVARAJ, Bahattin Koc, Stuart Maloy, Additive Manufacturing Vol. 51 2022 102601 In this study, a novel alloy of modified Inconel 718 produced by laser powder bed fusion is studied before and after in-situ Kr irradiation up to 3 dpa at 200 and 450 °C. Before irradiation, the microstructure consists of dislocation cells having a misorientation angle less than 5° and with an average size of ~500 nm. There are also second phase particles of MC type carbides, Laves phase and oxides such as Y-O, Y-(Ti)-Al-O. While the microstructure consists of stacking fault tetrahedra, faulted and perfect loops after irradiation at 200 °C, dislocation loops are the primary defects at 450 °C. With increasing dose, the size of the defects remains similar at 200 °C while it increases at 450 °C. This has been attributed to the existence of vacancy type defects at 200 °C and the different defect transport mechanisms at different temperatures. Moreover, matrix and second phase particle compositions seem to be similar after irradiation. The sink strengths of the structures have been calculated and superior radiation resistance of this alloy has been attributed to the existence of fine cell boundaries stabilized by the second phase particles produced by additive manufacturing.
		"Ion-irradiation-induced microstructural modifications in ferritic/martensitic steel T91" Xiang Liu, Yinbin Miao, Meimei Li, Marquis Kirk, Stuart Maloy, James Stubbins, Journal of Nuclear Materials Vol. 490 2017 305-316 Link In this paper, in situ transmission electron microscopy investigations were carried out to study the microstructural evolution of ferritic/martensitic steel T91 under 1 MeV Krypton ion irradiation up to 4.2 × 1015 ions/cm2 at 573 K, 673 K, and 773 K. At 573 K, grown-in defects are strongly modified by black-dot loops, and dislocation networks together with black-dot loops were observed after irradiation. At 673 K and 773 K, grown-in defects are only partially modified by dislocation loops; isolated loops and dislocation segments were commonly found after irradiation. Post irradiation examination indicates that at 4.2 × 1015 ions/cm2, about 51% of the loops were type for the 673 K irradiation, and the dominant loop type was for the 773 K irradiation. Finally, a dispersed barrier hardening model was employed to estimate the change in yield strength, and the calculated ion data were found to follow the similar trend as the existing neutron data with an offset of 100–150 MPa.
		"Kinetics of the Migration and Clustering of Extrinsic Gas in bcc Metals" Maria Okuniewski, James Stubbins, Chaitanya Deo, Srinivasan Srivilliputhur, Michael Baskes, Stuart Maloy, M. R. James, ASTM Special Technical Publication Vol. 1492 2008 177-189 Link We study the mechanisms by which gas atoms such as helium and hydrogen diffuse and interact with other defects in bcc metals and investigate the effect of these mechanisms on the nucleation of embryonic gas bubbles. Large quantities of helium and hydrogen are produced due to spallation and transmutation in structural materials in fusion and accelerator-driven reactors. The long time evolution of the extrinsic gas atoms and their accumulation at vacancies is studied using a kinetic Monte Carlo algorithm that is parameterized by the migration energies of the point defect entities. First-order reaction kinetics are observed when gas clusters with vacancies. If gas-gas clustering is allowed, mixed-order diffusion limited kinetics are observed. When dissociation of gas from clusters is allowed, gas-vacancy clusters survive to steady state while gas-gas clusters dissolve. We obtain cluster size distributions and reaction rate constants that can be used to quantify microstructural evolution of the irradiated metal.
		"Microstructure response of ferritic/martensitic steel HT9 after neutron irradiation: Effect of temperature" Ce Zheng, Elaina Reese, Kevin Field, Tian Liu, Emmanuelle Marquis, Stuart Maloy, Djamel Kaoumi, Journal of Nuclear Materials Vol. 528 2019 Link The ferritic/martensitic steel HT9 was irradiated in the BOR-60 reactor at 650, 690 and 730 K (377, 417 and 457 °C) to doses between ∼14.6–18.6 displacements per atom (dpa). Irradiated samples were comprehensively characterized using analytical scanning/transmission electron microscopy and atom probe tomography, with emphasis on the influence of irradiation temperature on microstructure evolution. Mn/Ni/Si-rich (G-phase) and Cr-rich (αʹ) precipitates were observed within martensitic laths and at various defect sinks at 650 and 690 K (377 and 417 °C). For both G-phase and αʹ precipitates, the number density decreased while the size increased with increasing temperature. At 730 K (457 °C), within martensitic laths, a very low density of large G-phase precipitates nucleating presumably on dislocation lines was observed. No αʹ precipitates were observed at this temperature. Both a <100> and a/2 <111> type dislocation loops were observed, with the a <100> type being the predominant type at 650 and 690 K (377 and 417 °C). On the contrary, very few dislocation loops were observed at 730 K (457 °C), and the microstructure was dominated by a/2 <111> type dislocation lines (i.e., dislocation network) at this temperature. Small cavities (diameter < 2 nm) were observed at all three temperatures, whereas large cavities (diameter > 2 nm) were observed only at 690 K (417 °C), resulting in a bimodal cavity size distribution at 690 K (417 °C) and a unimodal size distribution at 650 and 730 K (377 and 457 °C). The highest swelling (%) was observed at 690 K (417 °C), indicating that the peak of swelling happens between 650 and 730 K (377 and 457 °C).
		"Nanohardness measurements of heavy ion irradiated coarse- and nanocrystalline-grained tungsten at room and high temperature" Osman El Atwani, Jordan Weaver, JESUS ALFREDO ESQUIVEL, Yongqiang Wang, Stuart Maloy, Nathan Mara, Journal of Nuclear Materials Vol. 509 2018 276-284 Link
		"Neutron irradiation effects in Fe and Fe-Cr at 300°C" Wei-Ying Chen, Yinbin Miao, Jian Gan, Maria Okuniewski, Stuart Maloy, James Stubbins, Acta Materialia Vol. 111 2016 407-416 Link Fe and Fe-Cr (Cr = 10-16 at.%) specimens were neutron-irradiated at 300 °C to 0.01, 0.1 and 1 dpa. The TEM observations indicated that the Cr significantly reduced the mobility of dislocation loops and suppressed vacancy clustering, leading to distinct damage microstructures between Fe and Fe-Cr. Irradiation-induced dislocation loops in Fe were heterogeneously observed in the vicinity of grown-in dislocations, whereas the loop distribution observed in Fe-Cr is much more uniform. Voids were observed in the irradiated Fe samples, but not in irradiated Fe-Cr samples. Increasing Cr content in Fe-Cr results in a higher density, and a smaller size of irradiation-induced dislocation loops. Orowan mechanism was used to correlate the observed microstructure and hardening, which showed that the hardening in Fe-Cr can be attributed to the formation of dislocation loops and α′ precipitates.
		"Projection-capacitor discharge resistance welding of 430 stainless steel and 14YWT" Thomas Lienert, Calvin Lear, Todd Steckley, Lindsey Lindamood, Jerry Gould, Stuart Maloy, Ben Eftink, Journal of Manufacturing Processes Vol. 75 2022 1189-1201 Link Efforts to advance structural materials with improved properties and service life in support of next generation designs for nuclear reactor components have recently led to development of nano-ferritic alloys (NFAs) containing nano-oxides such as 14YWT. A key enabling technology to realizing the useful properties of NFAs during service involves preservation of the oxide dispersions during joining. Solid-state welding processes, such as projection-capacitor discharge resistance welding (P-CDRW) used here, are well suited for joining NFAs while retaining the oxides. Due to limitations in the supply of 14YWT NFA material, initial experiments were conducted using 430 stainless steel as an inexpensive surrogate material. The goal of the surrogate experiments was to scale suitable parameters from 430 welds to 14YWT using ratios of key properties for the two materials including flow stress at temperatures and strain rates relevant to hot working. Results indicated that weld displacement increased with increasing weld force and increasing weld energy for all other variables held constant. Weld energy appeared to have a larger effect on displacement than weld force for the sample geometry used here. Appropriate process parameters (no melting) were established for the two materials. The process window for the 430 material extended from 350 J to 600 J of energy for weld forces of 2.2 kN and 3.1 kN. Suitable parameters for 14YWT were similar in terms of energy but for force levels of 3.1 kN and 4.0 kN. Displacement for both materials ranged from 150 μm to 300 μm for welds that did not experience melting. Simple heat flow analysis confirmed that the extent of displacement was limited by the characteristic thermal distance determined from thermo-physical properties and the weld current rise time. The higher flow stresses of 14YWT relative to 430 were apparently offset by greater heating due to higher electrical resistivity near the projection tip and lesser heat conduction from the projection tip owing to lower thermal conductivity. Based on the results presented here and in our companion paper. The P-CDRW process appears capable of successfully joining the 14YWT NFA while retaining the microstructures and properties of the original material.
		"Radiation resistance of oxide dispersion strengthened alloys: Perspectives from in situ observations and rate theory calculations" Xiang Liu, Yinbin Miao, Meimei Li, Marquis Kirk, Guangming Zhang, Shigeharu Ukai, Stuart Maloy, James Stubbins, Scripta Materialia Vol. 148 2018 33-36 Link Here, in situ ion irradiation and rate theory calculations were employed to directly compare the radiation resistance of an oxide dispersion strengthened alloy with that of a conventional ferritic/martensitic alloy. Compared to the rapid buildup of dislocation loops, loop growth, and formation of network dislocations in the conventional ferritic/martensitic alloy, the superior radiation resistance of the oxide dispersion strengthened alloy is manifested by its stable dislocation structure under the same irradiation conditions. The results are consistent with rate theory calculations, which show that high-density nanoparticles can significantly reduce freely migrating defects and suppress the buildup of clustered defects.
		"Resilient ZnO nanowires in an irradiation enviroment: an in situ study" Cheng Sun, Jin Li, Youxing Chen, Mark Kirk, Meimei Li, Stuart Maloy, Haiyan Wang, Xinghang Zhang, Acta Materialia Vol. 95 2015 156 Link
		"Solid state welding of the nanostructured ferritic alloy 14YWT using a capacitive discharge resistance welding technique" Calvin Lear, Jonathan Gigax , Matt Schneider, Todd Steckley, Thomas Lienert, Stuart Maloy, Ben Eftink, Metals Vol. 12 2021 23 Link Joining nanostructured ferritic alloys (NFAs) has proved challenging, as the nano-oxides that provide superior strength, creep resistance, and radiation tolerance at high temperatures tend to agglomerate, redistribute, and coarsen during conventional fusion welding. In this study, capacitive discharge resistance welding (CDRW)—a solid-state variant of resistance welding—was used to join end caps and thin-walled cladding tubes of the NFA 14YWT. The resulting solid-state joints were found to be hermetically sealed and were characterized across the weld region using electron microscopy (macroscopic, microscopic, and nanometer scales) and nanoindentation. Microstructural evolution near the weld line was limited to narrow (~50–200 μm) thermo-mechanically affected zones (TMAZs) and to a reduction in pre-existing component textures. Dispersoid populations (i.e., nano-oxides and larger oxide particles) appeared unchanged by all but the highest energy and power CDRW condition, with this extreme producing only minor nano-oxide coarsening (~2 nm → ~5 nm Ø). Despite a minimal microstructural change, the TMAZs were found to be ~10% softer than the surrounding base material. These findings are considered in terms of past solid-state welding (SSW) efforts—cladding applications and NFA-like materials in particular—and in terms of strengthening mechanisms in NFAs and the potential impacts of localized temperature–strain conditions during SSW.
		"Stability of nanoclusters in an oxide dispersion strengthened alloy under neutron irradiation" Xiang Liu, Yinbin Miao, Yaqiao Wu, Stuart Maloy, James Stubbins, Scripta Materialia Vol. 138 2017 57-61 Link Here, we report atom probe tomography results of the nanoclusters in a neutron-irradiated oxide dispersion strengthened alloy. Following irradiation to 5 dpa at target temperatures of 300 °C and 450 °C, fewer large nanoclusters were found and the residual nanoclusters tend to reach an equilibrium Guinier radius of 1.8 nm. With increasing dose, evident decrease in peak oxygen and titanium (but not yttrium) concentrations in the nanoclusters was observed, which was explained by atomic weight, solubility, diffusivity, and chemical bonding arguments. The chemical modifications indicate the equilibrium size is indeed a balance of two competing processes: radiation enhanced diffusion and collisional dissolution.
		"Study of Irradiated Mod. 9Cr-1Mo Steel by Synchrotron XAS" Hasitha Ganegoda, Daniel Olive, Jeff Terry, Yulia Trenikhina, Meimei Li, Stuart Maloy, Transactions of the American Nuclear Society Vol. 102 2010 855 Link
		"Study of irradiated mod.9Cr–1Mo steel by synchrotron extended X-ray absorption fine structure" Hasitha Ganegoda, Meimei Li, Stuart Maloy, Jeff Terry, Yulia Trenikhina, Daniel Olive, Journal of Nuclear Materials Vol. 441 2013 674-680 Link Synchrotron extended X-ray absorption fine structure (EXAFS) spectroscopy measurements were performed to study the dose dependence of and alloying effects on irradiation-induced changes in the local atomic environments in a mod.9Cr–1Mo ferritic-martensitic steel. The measurements were carried out at room temperature on non-irradiated and irradiated specimens exposed to 1, 4, and 10 displacement per atom (dpa) at 40–70 °C. The EXAFS data for Fe, Cr, Mo, and Nb K-edges were recorded, and the local structure close to the X-ray absorbing atom was determined. Irradiation caused significant reductions in peak amplitude in the Fe, Mo and Nb K-edge Fourier transformed EXAFS. The data showed a systematic decrease in coordination number of neighbor atoms with increasing irradiation dose, and the dose dependence of the coordination loss was dependent on the specific element. The measured damage around Fe sites can be correlated with the dpa value, while the loss of near neighbors around Mo saturated at ~1 dpa. The coordination in the Fe matrix was reduced less by irradiation than either the coordination of Mo in solution or Nb in carbides. It was demonstrated that EXAFS can provide a detailed, atomic level description of radiation damage in complex alloy systems
		"The Accelerator Production of Tritium Materials Test Program" Stuart Maloy, Walter Sommer, Michael James, Tobias Romero, Manuel Lopez, Eugene Zimmermann, James Ledbetter, Nuclear Technology Vol. 132 2000 103-114 Link A materials qualification program has been developed to irradiate and test candidate materials (alloy 718, Type 316L, and Type 304L stainless steel, modified Fe9Cr-1Mo(T91), Al-6061-T6, and Al-5052-O) for use in the Accelerator Production of Tritium (APT) target and blanket. The irradiations were performed in prototypic proton and neutron spectra at prototypic temperatures (50 to 160°C). The study used the 800-MeV, 1.0-mA proton accelerator at the Los Alamos Neutron Science Center, which produces a Gaussian beam with 2 sigma = 3 cm. The experiment geometry is arranged to contain near-prototypic modules of the tungsten neutron source and the lead and aluminum blanket as well as mechanical test specimens of candidate APT materials. The particle spectrum varies throughout the irradiation volume; specimens are exposed to protons and a variety of mixed proton and neutron spectra, depending on the specimen’s position relative to the beam center. These specimens have been irradiated for >3600 h to a maximum proton fluence of 4 × 1021 p/cm2 in the center of the proton beam. Specimens will yield data on the effect of proton irradiation, to high dose, on material properties from tensile tests, three-point bend tests, fracture toughness tests, pressurized tubes, U-bend stress corrosion cracking specimens, corrosion measurements, and microstructural characterization using transmission electron microscopy specimens. Results from these studies are applicable to all spallation neutron sources now in operation and under consideration, including the Spallation Neutron Source, the European Spallation Source, and The Accelerator Transmutation of Waste project.

Presentations:

	"Atom probe analysis of a neutron irradiated Fe-14Cr model alloy" Jian Gan, Stuart Maloy, Yinbin Miao, Kun Mo, Yaqiao Wu, ICFRM 2013 January 1-9, (2013)
	"Capacitive Discharge Resistance Welding for ODS Steel Cladding: Weld Properties and Radiation Resistance" Calvin Lear, Ben Eftink, Thomas Lienert, Stuart Maloy, Todd Steckley, Materials in Nuclear Energy Systems (MiNES) October 6-10, (2019) Link
	"Capacitive Discharge Resistance Welding of 14YWT and Other Alloys" Calvin Lear, Ben Eftink, Lindsey Lindamood, Todd Steckley, Matt Schneider, Jerry Gould, Thomas Lienert, Stuart Maloy, TMS 2020 Annual Meeting & Exhibition February 23-27, (2020)
	"Characterization of Nanostructured Ferritic Alloy Atomized with Yttrium And Controlling Oxygen Content" Nicholas Cunningham, David Hoelzer, Stuart Maloy, G. Robert Odette, TMS 2014 February 16-20, (2014)
	"Development and Testing Advanced Ferritic Steels for Fast Reactor Applications" Osman Anderoglu, Stuart Maloy, G. Robert Odette, Tarik Saleh, TMS 2014 February 16-20, (2014)
	"Impact of Capacitive Discharge Resistance Welding on the Radiation Tolerance of 14YWT Cladding" Calvin Lear, Hyosim Kim, Matt Schneider, Todd Steckley, Yongqiang Wang, Thomas Lienert, Stuart Maloy, Ben Eftink, Structural Materials for Innovative Nuclear Systems (SMINS-6) September 12-15, (2022)
	"Manufacturing of nanostructured ODS steel cladding tubes for advanced nuclear reactors using cold spray technology" Mia Lenling, Hwasung Yeom, Ben Maier, Greg Johnson, David Hoelzer, Peter Hosemann, Stuart Maloy, Kumar Sridharan, Jeffrey Graham, The Minerals, Metals and Materials Society (TMS) 2019 March 10-14, (2019)
	"Microstructure and Mechanical Property Studies on Neutron-Irradiated Ferritic FeCr Model Alloys" Jian Gan, Stuart Maloy, Yinbin Miao, Kun Mo, James Stubbins, Yaqiao Wu, TMS Annual Meeting February 16-20, (2014)
	"Post Irradiation Examination of Fast Neutron Irradiated 14YWT Tubes at Nuclear Science User Facilities" Eda Aydogan, Peter Hosemann, David Krumwiede, Stuart Maloy, Tarik Saleh, 2017 ANS Annual Meeting [unknown]
	"Program Review Presentation Entitled "Development of Low Temperature Spray Process for Manufacturing Fuel Cladding and Surface Modification of Reactor Components"" Mia Lenling, Kumar Sridharan, Hwasung Yeom, Peter Hosemann, David Hoelzer, Jeffrey Graham, Stuart Maloy, Advanced Methods for Manufacturing Program Review December 4-6, (2018)

Accomplishments

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NSUF Supported Research

LANSCE APT Irradiation 1996/1997/1999 - Library Submission, #4613

NSUF Research Collaborations

3-D strain and phase mapping in AM Fe-9Cr steel - FY 2020 RTE 2nd Call, #4119

Atom probe characterization of HT-9 as a function of neutron irradiation temperature - FY 2023 RTE 2nd Call, #4629

Atom probe tomography on highly irradiated F/M steels - FY 2011 RTE Solicitation, #306

Characterization of the Microstructures and Mechanical Properties of Advanced Structural Alloys for Radiation Service: A Comprehensive Library of ATR Irradiated Alloys and Specimen - FY 2008 Call for User Proposals, #139

Effect of interstitial elements on the irradiation response of HT9 tempered ferritic/martensitic steels - FY 2019 RTE 3rd Call, #2883

Effects of carbon addition on the solute redistribution in Fe-9Cr alloys under irradiation - FY 2016 RTE 1st Call, #624

Elemental effects on radiation damage in tempered martensitic steels neutron irradiated to high doses at fast reactor relevant temperatures - FY 2024 CINR, #5020

In-Situ SEM Irradiation Enhanced Creep Studies of 14 YWT - FY 2019 RTE 1st Call, #1648

Investigating nitrogen effects on the mechanical properties and microstructure evolution in neutron irradiated HT-9 steel - FY 2023 RTE 2nd Call, #4651

Investigating the Performance of Refractory High Entropy Alloys Under Irradiation and Mechanical Extremes - FY 2020 RTE 1st Call, #3026

Irradiation Performance of Fe-Cr Base Alloys - FY 2008 Call for User Proposals, #92

Nanohardness Measurements on Neutron Irradiated Steel Samples for Next Generation Reactors3 - FY 2015 RTE 3rd Call, #585

Performance of Nanocrystalline and Ultrafine Tungsten Under Irradiation and Mechanical Extremes - FY 2017 RTE 2nd Call, #951

Room Temperature Tensile Properties of ATR Neutron Irradiated T91 - FY 2023 RTE 2nd Call, #4705

Synchrotron X-ray Diffraction Study of Microstructural Evolution in Irradiated Mod.9Cr-1Mo Steel - FY 2011 APS, #264

Understanding fundamental effect of grain structure on microstructure evolution in HT9 via in-situ irradiation and TEM - FY 2023 RTE 3rd Call, #4744