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Now showing items 1 - 16 of 4500

  • Chemical vapor deposition of Mo tubes for fuel cladding applications

    Miles F. Beaux   Douglas R. Vodnik   Reuben J. Peterson   Bryan L. Bennett   Jesse J. Salazar   Terry G. Holesinger   Graham King   Stuart A. Maloy   David J. Devlin   Igor O. Usov  

    Abstract Chemical vapor deposition (CVD) techniques have been evaluated for fabrication of free-standing 0.25 mm thick molybdenum tubes with the end goal of nuclear fuel cladding applications. In order to produce tubes with the wall thickness and microstructures desirable for this application, long deposition durations on the order of 50 h with slow deposition rates were employed. A standard CVD method, involving molybdenum pentachloride reduction by hydrogen, as well as a fluidized-bed CVD (FBCVD) method was applied towards these objectives. Characterization of the tubes produced in this manner revealed regions of material with fine grain microstructure and wall thickness suitable for fuel cladding applications, but lacking necessary uniformity across the length of the tubes. A path forward for the production of freestanding molybdenum tubes that possess the desired properties across their entire length has been identified and can be accomplished by future optimization of the deposition system. Highlights • Free-standing molybdenum tubes produced by chemical vapor deposition techniques. • Properties suitable for use of tubes as nuclear fuel cladding were achieved locally. • Proof of concept for production of suitable tubes has been demonstrated. • Composition, microstructure, and phase characterized by ICP, XRF, SEM and XRD.
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  • Ion-irradiation-induced microstructural modifications in ferritic/martensitic steel T91

    Xiang Liu   Yinbin Miao   Meimei Li   Marquis A. Kirk   Stuart A. Maloy   James F. Stubbins  

    Abstract 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 × 10 15 ions/cm 2 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 × 10 15 ions/cm 2 , about 51% of the loops were a 0 / 2 〈 111 〉 type for the 673 K irradiation, and the dominant loop type was a 0 〈 100 〉 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. Graphical abstract Image 1 • Microstructure evolution of T91 under Krypton ion irradiation was studied. • Distinct microstructure evolution was explained by 〈111>−<100〉 loop transition. • A hardening model was used to establish the microstructure-property relationship.
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  • Tensile testing of EP-823 and HT-9 after irradiation in STIP II

    Stuart A. Maloy   T. Romero   M.R. James   Y. Dai  

    The predicted operating conditions for a lead-bismuth eutectic target to be used in an accelerator driven system for the Advanced Fuel Cycle Initiative spans a temperature range of 300-600degC while being irradiated by a high energy (~600MeV) proton beam. Such spallation conditions lead to high displacement rates coupled with high accumulation rates of helium and hydrogen up to 150appm/dpa. Two candidate ferritic/martensitic materials for these applications include HT-9 and EP-823. To investigate the effect of irradiation on these materials, the tensile properties were measured at 25, 250, and 400degC after irradiation in the STIP II irradiation to doses up to 20dpa at temperatures up to 350degC. Losses in ductility concomitant with increases in yield stress are observed in both alloys although the embrittlement is more severe in the EP-823. This stronger embrittlement is attributed to the high silicon content in EP-823. [All rights reserved Elsevier]
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  • Tensile testing of EP-823 and HT-9 after irradiation in STIP II

    Stuart A. Maloy   T. Romero   M.R. James and Y. Dai  

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  • Preface

    Stuart A. Maloy   Christina Trautmann   Gary S. Was  

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  • Preface

    Stuart A. Maloy   Christina Trautmann   Gary S. Was  

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  • Thermal annealing recovery of fracture toughness in HT9 steel after irradiation to high doses

    Thak Sang Byun   Jong-Hyuk Baek   Osman Anderoglu   Stuart A. Maloy   Mychailo B. Toloczko  

    Abstract The HT9 ferritic/martensitic steel with a nominal chemistry of Fe(bal.)12%Cr1%MoVW has been used as a primary core material for fast fission reactors such as FFTF because of its high resistance to radiation-induced swelling and embrittlement. Both static and dynamic fracture test results have shown that the HT9 steel can become brittle when it is exposed to high dose irradiation at a relatively low temperature (<430 °C). This article aims at a comprehensive discussion on the thermal annealing recovery of fracture toughness in the HT9 steel after irradiation up to 3148 dpa at 378504 °C. A specimen reuse technique has been established and applied to this study: the fracture specimens were tested Charpy specimens or broken halves of Charpy bars (13 × 3 × 4 mm). The post-anneal fracture test results indicated that much of the radiation-induced damage can be recovered by a simple thermal annealing schedule: the fracture toughness was incompletely recovered by 550 °C annealing, while nearly complete or complete recovery occurred after 650 °C annealing. This indicates that thermal annealing is a feasible damage mitigation technique for the reactor components made of HT9 steel. The partial recovery is probably due to the non-removable microstructural damages such as void or gas bubble formation, elemental segregation and precipitation.
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  • Water corrosion measurements on tungsten irradiated with high energy protons and spallation neutrons

    Stuart A. Maloy   R. Scott Lillard   Walter F. Sommer   Darryl P. Butt   Frank D. Gac   Gordon J. Willcutt   McIntyre R. Louthan Jr.  

    A detailed analysis was performed on the degradation of a tungsten target under water cooling while being exposed to a 761MeV proton beam at an average current of 0.867mA to a maximum fluence of 1.3times10 21 protons/cm 2. The target consisted of 3mm diameter tungsten rods arranged in bundles and cooled with deionized water flowing over their length. Degradation of the tungsten was measured through analyzing water resistivity, tungsten concentration in water samples that were taken during irradiation and through dimensional measurements on the rods after irradiation. Chemical analysis of irradiated water samples showed W concentrations up to 35mug/ml. Gamma analysis showed increases in concentrations of many isotopes including W-178, Lu-171, Tm-167, Tm-166, Yb-169 and Hf-175. Dimensional measurements performed after irradiation on the W rods revealed a decrease in diameter as a function of position that followed closely the Gaussian proton beam profile along the rod length and indicated a definite beam-effect. A general decrease in diameter, especially on the coolant-water entrance point where turbulent flow was likely, also suggests a chemically and mechanically-driven corrosion effect. A method to estimate the apparent corrosion rate based on proton fluence is presented and application of this method estimates the material loss rate at about 1.9W atoms/incident proton. From this result, the corrosion rate of tungsten in a 761MeV, 0.867mA proton beam was calculated to be 0.073cm/full power year. of irradiation. [All rights reserved Elsevier].
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  • Atom probe study of irradiation-enhanced α′ precipitation in neutron-irradiated Fe–Cr model alloys

    Wei-Ying Chen   Yinbin Miao   Yaqiao Wu   Carolyn A. Tomchik   Kun Mo   Jian Gan   Maria A. Okuniewski   Stuart A. Maloy   James F. Stubbins  

    Abstract 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 × 10 23 to 2.7 × 10 24 1/m 3 , 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.
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  • High strain rate deformation of NiAl

    Stuart A. Maloy   George T. Gray III   Ram Darolia  

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  • The high temperature three point bend testing of proton irradiated 316L stainless steel and Mod 9Cr–1Mo

    Stuart A. Maloy   A. Zubelewicz   T. Romero   M.R. James   W.F. Sommer   Y. Dai  

    The predicted operating conditions for a lead-bismuth eutectic target to be used in an accelerator driven system for the Advanced Fuel Cycle Initiative span a temperature range of 300-600degC while being irradiated by a high energy (~600MeV) proton beam. Such spallation conditions lead to high displacement rates coupled with high accumulation rates of helium and hydrogen up to 150appm/dpa. Some candidate materials for these applications include Mod9Cr-1Mo and 316L stainless steel. To investigate the effect of irradiation on these materials, the mechanical properties are being measured through three point bend testing on Mod 9Cr-1Mo and 316L at 25, 250, 350 and 500degC after irradiation in a high energy proton beam (500-800MeV) to a dose of 9.8dpa at temperatures from 200 to 320degC. By comparing measurements made in bending to tensile measurements measured on identically irradiated materials, a measurement of 0.2% offset yield stress was obtained from 0.05% offset yield stress measured in three point bend testing. Yield stress increased by more than a factor of two after irradiation to 9.8dpa. Observation of the outer fiber surface of 316L showed very localized deformation when tested after irradiation at 70degC and deformation on multiple slip systems when tested after irradiation at 250-320degC. [All rights reserved Elsevier]
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  • The effects of fast reactor irradiation conditions on the tensile properties of two ferritic/martensitic steels

    Stuart A. Maloy   M.B. Toloczko   K.J. McClellan   T. Romero   Y. Kohno   F.A. Garner   R.J. Kurtz   A. Kimura  

    Tensile testing has been performed at 25 and at ~400degC on two ferritic/martensitic steels (JFMS and HT-9) after irradiation in FFTF to up to ~70dpa at 373-433degC. As observed in previous studies, this range of irradiation temperatures has a significant effect on hardening. The percent increase in yield stress decreases with increasing irradiation temperature from 373 to 433degC. The JFMS alloy, which has 0.7wt% silicon, exhibits approximately a factor of two increase in yield strength between tests at 427 and at 373degC, and shows an increase in hardening with increasing dose. A comparison of the JFMS tensile properties to the properties of other ferritic/martensitic steels suggests that this hardening is due to precipitation of a Si-rich laves phase in this alloy. The HT-9 alloy, which contains more chromium and more carbon but less silicon (0.2wt%), less molybdenum and less nickel, hardens during irradiation at 373degC, but shows less hardening for irradiations performed at 427degC and no increase in yield stress with increasing dose beyond 10dpa. [All rights reserved Elsevier]
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  • The effects of fast reactor irradiation conditions on the tensile properties of two ferritic/martensitic steels

    Stuart A. Maloy   M.B. Toloczko   K.J. McClellan   T. Romero   Y. Kohno   F.A. Garner   R.J. Kurtz and A. Kimura  

    Tensile testing has been performed at 25 and at 400 °C on two ferritic/martensitic steels (JFMS and HT-9) after irradiation in FFTF to up to 70 dpa at 373–433 °C. As observed in previous studies, this range of irradiation temperatures has a significant effect on hardening. The percent increase in yield stress decreases with increasing irradiation temperature from 373 to 433 °C. The JFMS alloy, which has 0.7 wt%silicon, exhibits approximately a factor of two increase in yield strength between tests at 427 and at 373 °C, and shows an increase in hardening with increasing dose. A comparison of the JFMS tensile properties to the properties of other ferritic/martensitic steels suggests that this hardening is due to precipitation of a Si-rich laves phase in this alloy. The HT-9 alloy, which contains more chromium and more carbon but less silicon (0.2 wt%), less molybdenum and less nickel, hardens during irradiation at 373 °C, but shows less hardening for irradiations performed at 427 °C and no increase in yield stress with increasing dose beyond 10 dpa.
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  • A comparative assessment of the fracture toughness behavior of ferritic-martensitic steels and nanostructured ferritic alloys

    Thak Sang Byun   David T. Hoelzer   Jeoung Han Kim   Stuart A. Maloy  

    Abstract The Fe-Cr alloys with ultrafine microstructures are primary candidate materials for advanced nuclear reactor components because of their excellent high temperature strength and high resistance to radiation-induced damage such as embrittlement and swelling. Mainly two types of Fe-Cr alloys have been developed for the high temperature reactor applications: the quenched and tempered ferritic-martensitic (FM) steels hardened primarily by ultrafine laths and carbonitrides and the powder metallurgy-based nanostructured ferritic alloys (NFAs) by nanograin structure and nanoclusters. This study aims at elucidating the differences and similarities in the temperature and strength dependences of fracture toughness in the Fe-Cr alloys to provide a comparative assessment of their high-temperature structural performance. The K JQ versus yield stress plots confirmed that the fracture toughness was inversely proportional to yield strength. It was found, however, that the toughness data for some NFAs were outside the band of the integrated dataset at given strength level, which indicates either a significant improvement or deterioration in mechanical properties due to fundamental changes in deformation and fracture mechanisms. When compared to the behavior of NFAs, the FM steels have shown much less strength dependence and formed narrow fracture toughness data bands at a significantly lower strength region. It appeared that at high temperatures ≥600 °C the NFAs cannot retain the nanostructure advantage of high strength and high toughness either by high-temperature embrittlement or by excessive loss of strength. Irradiation studies have revealed, however, that the NFAs have much stronger radiation resistance than tempered martensitic steels, such as lower radiation-induced swelling, finer helium bubble formation, lower irradiation creep rate and reduced low temperature embrittlement.
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  • Effects of carbon additions on the high temperature mechanical properties of molybdenum disilicide

    Stuart A. Maloy   John J. Lewandowski   Arthur H. Heuer   John J. Petrovic  

    Additions of up to 4 wt.% C to MoSi 2 caused elimination of the siliceous grain boundary phase in hot pressed samples as well as the formation of small quantities of beta-SiC and Mo 5Si 3C. Both the hardness and fracture toughness of the carbon-containing alloys exceeded those of the carbon-free (and oxygen-rich) materials. The toughness for all carbon-containing alloys at 800-1400degC was independent of the level of carbon addition, although the hardness at all temperatures up to 1000degC increased with the level of carbon addition. Unstable fracture was exhibited by all alloys tested below 1200degC, while stable cracking and high toughness (greater than or equal to 11.5 MPa m 1/2) was obtained in the 2 wt.% C alloy at 1400degC
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  • A conversation with Elizabeth A. Stuart

    Sherri Rose  

    Elizabeth A. Stuart is a Professor in the Departments of Mental Health, Biostatistics, and Health Policy and Management at the Johns Hopkins Bloomberg School of Public Health, and Associate Dean for Education at the school. She is a renowned expert in the area of causal inference, including propensity score methods for observational data and the generalizability of randomized trial results, and is also a Fellow of the American Statistical Association. Prior to her appointment to the faculty at Johns Hopkins, Professor Stuart received her Ph.D. in statistics from Harvard University and a bachelor’s degree in mathematics from Smith College. In 2015, Professor Stuart was recognized at the International Conference on Health Policy Statistics with the Mid-Career Award from the Health Policy Statistics Section of the American Statistical Association.
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