Preliminary investigations on the use of uranium silicide targets for fission Mo-99 production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cols, H.; Cristini, P.; Marques, R.

1997-08-01

The National Atomic Energy Commission (CNEA) of Argentine Republic owns and operates an installation for production of molybdenum-99 from fission products since 1985, and, since 1991, covers the whole national demand of this nuclide, carrying out a program of weekly productions, achieving an average activity of 13 terabecquerel per week. At present they are finishing an enlargement of the production plant that will allow an increase in the volume of production to about one hundred of terabecquerel. Irradiation targets are uranium/aluminium alloy with 90% enriched uranium with aluminium cladding. In view of international trends held at present for replacing highmore » enrichment uranium (HEU) for enrichment values lower than 20 % (LEU), since 1990 the authors are in contact with the RERTR program, beginning with tests to adapt their separation process to new irradiation target conditions. Uranium silicide (U{sub 3}Si{sub 2}) was chosen as the testing material, because it has an uranium mass per volume unit, so that it allows to reduce enrichment to a value of 20%. CNEA has the technology for manufacturing miniplates of uranium silicide for their purposes. In this way, equivalent amounts of Molybdenum-99 could be obtained with no substantial changes in target parameters and irradiation conditions established for the current process with Al/U alloy. This paper shows results achieved on the use of this new target.« less

Studies on separation and purification of fission (99)Mo from neutron activated uranium aluminum alloy.

PubMed

Rao, Ankita; Kumar Sharma, Abhishek; Kumar, Pradeep; Charyulu, M M; Tomar, B S; Ramakumar, K L

2014-07-01

A new method has been developed for separation and purification of fission (99)Mo from neutron activated uranium-aluminum alloy. Alkali dissolution of the irradiated target (100mg) results in aluminum along with (99)Mo and a few fission products passing into solution, while most of the fission products, activation products and uranium remain undissolved. Subsequent purification steps involve precipitation of aluminum as Al(OH)3, iodine as AgI/AgIO3 and molybdenum as Mo-α-benzoin oxime. Ruthenium is separated by volatilization as RuO4 and final purification of (99)Mo was carried out using anion exchange method. The radiochemical yield of fission (99)Mo was found to be >80% and the purity of the product was in conformity with the international pharmacopoeia standards. Copyright © 2014 Elsevier Ltd. All rights reserved.

TERNARY ALLOY-CONTAINING PLUTONIUM

DOEpatents

Waber, J.T.

1960-02-23

Ternary alloys of uranium and plutonium containing as the third element either molybdenum or zirconium are reported. Such alloys are particularly useful as reactor fuels in fast breeder reactors. The alloy contains from 2 to 25 at.% of molybdenum or zirconium, the balance being a combination of uranium and plutonium in the ratio of from 1 to 9 atoms of uranlum for each atom of plutonium. These alloys are prepared by melting the constituent elements, treating them at an elevated temperature for homogenization, and cooling them to room temperature, the rate of cooling varying with the oomposition and the desired phase structure. The preferred embodiment contains 12 to 25 at.% of molybdenum and is treated by quenching to obtain a body centered cubic crystal structure. The most important advantage of these alloys over prior binary alloys of both plutonium and uranium is the lack of cracking during casting and their ready machinability.

Microstructures and Hardness/Wear Performance of High-Carbon Stellite Alloys Containing Molybdenum

NASA Astrophysics Data System (ADS)

Liu, Rong; Yao, J. H.; Zhang, Q. L.; Yao, M. X.; Collier, Rachel

2015-12-01

Conventional high-carbon Stellite alloys contain a certain amount of tungsten which mainly serves to provide strengthening to the solid solution matrix. These alloys are designed for combating severe wear. High-carbon molybdenum-containing Stellite alloys are newly developed 700 series of Stellite family, with molybdenum replacing tungsten, which are particularly employed in severe wear condition with corrosion also involved. Three high-carbon Stellite alloys, designated as Stellite 706, Stellite 712, and Stellite 720, with different carbon and molybdenum contents, are studied experimentally in this research, focusing on microstructure and phases, hardness, and wear resistance, using SEM/EDX/XRD techniques, a Rockwell hardness tester, and a pin-on-disk tribometer. It is found that both carbon and molybdenum contents influence the microstructures of these alloys significantly. The former determines the volume fraction of carbides in the alloys, and the latter governs the amount of molybdenum-rich carbides precipitated in the alloys. The hardness and wear resistance of these alloys are increased with the carbide volume fraction. However, with the same or similar carbon content, high-carbon CoCrMo Stellite alloys exhibit worse wear resistance than high-carbon CoCrW Stellite alloys.

Pre-treatment for molybdenum or molybdenum-rich alloy articles to be plated

DOEpatents

Wright, Ralph R.

1980-01-01

This invention is a method for etching a molybdenum or molybdenum-rich alloy surface to promote the formation of an adherent bond with a subsequently deposited metallic plating. In a typical application, the method is used as a pre-treatment for surfaces to be electrolessly plated with nickel. The pre-treatment comprises exposing the crystal boundaries of the surface by (a) anodizing the surface in acidic solution to form a continuous film of gray molybdenum oxide thereon and (b) removing the film.

THE GRAVIMETRIC DETERMINATION OF MOLYBDENUM IN URANIUM-MOLYBDENUM ALLOYS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1959-03-01

The sample is dissolved in nitric and hydrochloric acids. After heating the solution with sulfuric acid, molybodenum is precipitated as the benzoin-oxime complex which is ignited to molybdic oxide. This is dissolved in ammonia, and the molybdenum is precipitated and weighed as lead molybdate. (auth)

High-strength, creep-resistant molybdenum alloy and process for producing the same

DOEpatents

Bianco, Robert; Buckman, Jr., R. William; Geller, Clint B.

1999-01-01

A wet-doping process for producing an oxide-dispersion strengthened (ODS), creep-resistant molybdenum alloy is disclosed. The alloy is made by adding nitrate or acetate salts of lanthanum, cerium, thorium, or yttrium to molybdenum oxide to produce a slurry, heating the slurry in a hydrogen atmosphere to produce a powder, mixing and cold isostatically pressing the powder, sintering in a hydrogen atmosphere, and thermomechanically processing (swaging, extruding, cold drawing) the product. The ODS molybdenum alloy produced by the process contains 2-4% by volume (.about.1-4% by weight) of an oxide of lanthanum, cerium, thorium, or yttrium. The alloy has high strength and improved creep-resistance at temperatures greater than 0.55T.sub.m of molybdenum.

Process for electroslag refining of uranium and uranium alloys

DOEpatents

Lewis, P.S. Jr.; Agee, W.A.; Bullock, J.S. IV; Condon, J.B.

1975-07-22

A process is described for electroslag refining of uranium and uranium alloys wherein molten uranium and uranium alloys are melted in a molten layer of a fluoride slag containing up to about 8 weight percent calcium metal. The calcium metal reduces oxides in the uranium and uranium alloys to provide them with an oxygen content of less than 100 parts per million. (auth)

Role of electron concentration in softening and hardening of ternary molybdenum alloys

NASA Technical Reports Server (NTRS)

Stephens, J. R.; Witzke, W. R.

1975-01-01

Effects of various combinations of hafnium, tantalum, rhenium, osmium, iridium, and platinum in ternary molybdenum alloys on alloy softening and hardening were determined. Hardness tests were conducted at four test temperatures over the temperature range 77 to 411 K. Results showed that hardness data for ternary molybdenum alloys could be correlated with anticipated results from binary data based upon expressions involving the number of s and d electrons contributed by the solute elements. The correlation indicated that electron concentration plays a dominant role in controlling the hardness of ternary molybdenum alloys.

High-strength, creep-resistant molybdenum alloy and process for producing the same

DOEpatents

Bianco, R.; Buckman, R.W. Jr.; Geller, C.B.

1999-02-09

A wet-doping process for producing an oxide-dispersion strengthened (ODS), creep-resistant molybdenum alloy is disclosed. The alloy is made by adding nitrate or acetate salts of lanthanum, cerium, thorium, or yttrium to molybdenum oxide to produce a slurry, heating the slurry in a hydrogen atmosphere to produce a powder, mixing and cold isostatically pressing the powder, sintering in a hydrogen atmosphere, and thermomechanically processing (swaging, extruding, cold drawing) the product. The ODS molybdenum alloy produced by the process contains 2--4% by volume (ca. 1--4% by weight) of an oxide of lanthanum, cerium, thorium, or yttrium. The alloy has high strength and improved creep-resistance at temperatures greater than 0.55T{sub m} of molybdenum. 10 figs.

PREPARATION OF URANIUM-ALUMINUM ALLOYS

DOEpatents

Moore, R.H.

1962-09-01

A process is given for preparing uranium--aluminum alloys from a solution of uranium halide in an about equimolar molten alkali metal halide-- aluminum halide mixture and excess aluminum. The uranium halide is reduced and the uranium is alloyed with the excess aluminum. The alloy and salt are separated from each other. (AEC)

Spall fracture and strength of uranium, plutonium and their alloys under shock wave loading

NASA Astrophysics Data System (ADS)

Golubev, Vladimir

2015-06-01

Numerous results on studying the spall fracture phenomenon of uranium, two its alloys with molybdenum and zirconium, plutonium and its alloy with gallium under shock wave loading are presented in the paper. The majority of tests were conducted with the samples in the form of disks 4mm in thickness. They were loaded by the impact of aluminum plates 4mm thick through a copper screen serving as the cover or bottom part of a special container. The initial temperature of samples was changed in the range of -196 - 800 C degree for uranium and 40 - 315 C degree for plutonium. The character of spall failure of materials and the degree of damage for all tested samples were observed on the longitudinal metallographic sections of recovered samples. For a concrete test temperature, the impact velocity was sequentially changed and therefore the loading conditions corresponding to the consecutive transition from microdamage nucleation up to complete macroscopic spall fracture were determined. Numerical calculations of the conditions of shock wave loading and spall fracture of samples were performed in the elastoplastic approach. Several two- and three-dimensional effects of loading were taken into account. Some results obtained under conditions of intensive impulse irradiation and intensive explosive loading are presented too. The rather complete analysis and comparison of obtained results with the data of other researchers on the spall fracture of examined materials were conducted.

Method for fabricating uranium foils and uranium alloy foils

DOEpatents

Hofman, Gerard L [Downers Grove, IL; Meyer, Mitchell K [Idaho Falls, ID; Knighton, Gaven C [Moore, ID; Clark, Curtis R [Idaho Falls, ID

2006-09-05

A method of producing thin foils of uranium or an alloy. The uranium or alloy is cast as a plate or sheet having a thickness less than about 5 mm and thereafter cold rolled in one or more passes at substantially ambient temperatures until the uranium or alloy thereof is in the shape of a foil having a thickness less than about 1.0 mm. The uranium alloy includes one or more of Zr, Nb, Mo, Cr, Fe, Si, Ni, Cu or Al.

High strength uranium-tungsten alloys

DOEpatents

Dunn, Paul S.; Sheinberg, Haskell; Hogan, Billy M.; Lewis, Homer D.; Dickinson, James M.

1991-01-01

Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

Strength and fracture of uranium, plutonium and several their alloys under shock wave loading

NASA Astrophysics Data System (ADS)

Golubev, V. K.

2012-08-01

Results on studying the spall fracture of uranium, plutonium and several their alloys under shock wave loading are presented in the paper. The problems of influence of initial temperature in a range of - 196 - 800∘C and loading time on the spall strength and failure character of uranium and two its alloys with molybdenum and both molybdenum and zirconium were studied. The results for plutonium and its alloy with gallium were obtained at a normal temperature and in a temperature range of 40-315∘C, respectively. The majority of tests were conducted with the samples in the form of disks 4 mm in thickness. They were loaded by the impact of aluminum plates 4 mm thick through a copper screen 12 mm thick serving as the cover or bottom part of a special container. The character of spall failure of materials and the damage degree of samples were observed on the longitudinal metallographic sections of recovered samples. For a concrete test temperature, the impact velocity was sequentially changed and therefore the loading conditions corresponding to the consecutive transition from microdamage nucleation up to complete macroscopic spall fracture were determined. The conditions of shock wave loading were calculated using an elastic-plastic computer program. The comparison of obtained results with the data of other researchers on the spall fracture of examined materials was conducted.

Nuclear fuel alloys or mixtures and method of making thereof

DOEpatents

Mariani, Robert Dominick; Porter, Douglas Lloyd

2016-04-05

Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

PLUTONIUM-URANIUM-TITANIUM ALLOYS

DOEpatents

Coffinberry, A.S.

1959-07-28

A plutonium-uranium alloy suitable for use as the fuel element in a fast breeder reactor is described. The alloy contains from 15 to 60 at.% titanium with the remainder uranium and plutonium in a specific ratio, thereby limiting the undesirable zeta phase and rendering the alloy relatively resistant to corrosion and giving it the essential characteristic of good mechanical workability.

Molybdenum-copper and tungsten-copper alloys and method of making

DOEpatents

Schmidt, Frederick A.; Verhoeven, John D.; Gibson, Edwin D.

1989-05-23

Molybdenum-copper and tungsten-copper alloys are prepared by a consumable electrode method in which the electrode consists of a copper matrix with embedded strips of refractory molybdenum or tungsten. The electrode is progressively melted at its lower end with a superatmospheric inert gas pressure maintained around the liquifying electrode. The inert gas pressure is sufficiently above the vapor pressure of copper at the liquidus temperature of the alloy being formed to suppress boiling of liquid copper.

Molybdenum-copper and tungsten-copper alloys and method of making

DOEpatents

Schmidt, F.A.; Verhoeven, J.D.; Gibson, E.D.

1989-05-23

Molybdenum-copper and tungsten-copper alloys are prepared by a consumable electrode method in which the electrode consists of a copper matrix with embedded strips of refractory molybdenum or tungsten. The electrode is progressively melted at its lower end with a superatmospheric inert gas pressure maintained around the liquefying electrode. The inert gas pressure is sufficiently above the vapor pressure of copper at the liquidus temperature of the alloy being formed to suppress boiling of liquid copper. 6 figs.

High strength uranium-tungsten alloy process

DOEpatents

Dunn, Paul S.; Sheinberg, Haskell; Hogan, Billy M.; Lewis, Homer D.; Dickinson, James M.

1990-01-01

Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

Molybdenum disilicide alloy matrix composite

DOEpatents

Petrovic, John J.; Honnell, Richard E.; Gibbs, W. Scott

1990-01-01

Compositions of matter consisting of matrix matrials having silicon carbide dispersed throughout them and methods of making the compositions. A matrix material is an alloy of an intermetallic compound, molybdenum disilicide, and at least one secondary component which is a refractory silicide. The silicon carbide dispersant may be in the form of VLS whiskers, VS whiskers, or submicron powder or a mixture of these forms.

Molybdenum disilicide alloy matrix composite

DOEpatents

Petrovic, John J.; Honnell, Richard E.; Gibbs, W. Scott

1991-01-01

Compositions of matter consisting of matrix materials having silicon carbide dispersed throughout them and methods of making the compositions. A matrix material is an alloy of an intermetallic compound, molybdenum disilicide, and at least one secondary component which is a refractory silicide. The silicon carbide dispersant may be in the form of VLS whiskers, VS whiskers, or submicron powder or a mixture of these forms.

Acidic ammonothermal growth of gallium nitride in a liner-free molybdenum alloy autoclave

NASA Astrophysics Data System (ADS)

Malkowski, Thomas F.; Pimputkar, Siddha; Speck, James S.; DenBaars, Steven P.; Nakamura, Shuji

2016-12-01

This paper discusses promising materials for use as internal, non-load bearing components as well as molybdenum-based alloys for autoclave structural components for an ammonothermal autoclave. An autoclave was constructed from the commercial titanium-zirconium-molybdenum (TZM) alloy and was found to be chemically inert and mechanically stable under acidic ammonothermal conditions. Preliminary seeded growth of GaN was demonstrated with negligible incorporation of transition metals (including molybdenum) into the grown material (<1017 cm-3). Molybdenum and TZM were exposed to a basic ammonothermal environment, leading to slight degradation through formation of molybdenum nitride powders on their surface at elevated temperatures (T>560 °C). The possibility of a 'universal', inexpensive, liner-free ammonothermal autoclave capable of exposure to basic and acidic chemistry is demonstrated.

Molybdenum disilicide alloy matrix composite

DOEpatents

Petrovic, J.J.; Honnell, R.E.; Gibbs, W.S.

1991-12-03

Compositions of matter consisting of matrix materials having silicon carbide dispersed throughout them and methods of making the compositions are disclosed. A matrix material is an alloy of an intermetallic compound, molybdenum disilicide, and at least one secondary component which is a refractory silicide. The silicon carbide dispersant may be in the form of VLS whiskers, VS whiskers, or submicron powder or a mixture of these forms. 3 figures.

High strength and density tungsten-uranium alloys

DOEpatents

Sheinberg, Haskell

1993-01-01

Alloys of tungsten and uranium and a method for making the alloys. The amount of tungsten present in the alloys is from about 55 vol % to about 85 vol %. A porous preform is made by sintering consolidated tungsten powder. The preform is impregnated with molten uranium such that (1) uranium fills the pores of the preform to form uranium in a tungsten matrix or (2) uranium dissolves portions of the preform to form a continuous uranium phase containing tungsten particles.

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

DOEpatents

Travelli, A.

1985-10-25

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

DOEpatents

Travelli, Armando

1988-01-01

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Process for continuous production of metallic uranium and uranium alloys

DOEpatents

Hayden, H.W. Jr.; Horton, J.A.; Elliott, G.R.B.

1995-06-06

A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO{sub 3}), or any other substantially stable uranium oxide, to form the uranium dioxide (UO{sub 2}). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl{sub 4}), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation. 4 figs.

Process for continuous production of metallic uranium and uranium alloys

DOEpatents

Hayden, Jr., Howard W.; Horton, James A.; Elliott, Guy R. B.

1995-01-01

A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.

Process for alloying uranium and niobium

DOEpatents

Holcombe, Cressie E.; Northcutt, Jr., Walter G.; Masters, David R.; Chapman, Lloyd R.

1991-01-01

Alloys such as U-6Nb are prepared by forming a stacked sandwich array of uraniun sheets and niobium powder disposed in layers between the sheets, heating the array in a vacuum induction melting furnace to a temperature such as to melt the uranium, holding the resulting mixture at a temperature above the melting point of uranium until the niobium dissolves in the uranium, and casting the uranium-niobium solution. Compositional uniformity in the alloy product is enabled by use of the sandwich structure of uranium sheets and niobium powder.

Evaluation of Oxide Dispersion Strengthened (ODS) Molybdenum Alloys

DTIC Science & Technology

1997-01-01

rrSÄSTSÄ approximately 3900° E. Tungsten , molybdenum, »’^^^eÄfon^^Ä^Setttese techniques-are excellent candidates tor <^*Jf?£L5*!s3J to form oxides. The...1% creep strain in 1,000 hours) of thoriated tungsten alloys was measured to be up to five times higher than commercially-pure tungsten . These alloys...temperature decomposable hydroxide or carbonate oxide compound are mixed, Reference (d). The resulting powder batch mixture is then cold isostatically

Method of fabricating a uranium-bearing foil

DOEpatents

Gooch, Jackie G [Seymour, TN; DeMint, Amy L [Kingston, TN

2012-04-24

Methods of fabricating a uranium-bearing foil are described. The foil may be substantially pure uranium, or may be a uranium alloy such as a uranium-molybdenum alloy. The method typically includes a series of hot rolling operations on a cast plate material to form a thin sheet. These hot rolling operations are typically performed using a process where each pass reduces the thickness of the plate by a substantially constant percentage. The sheet is typically then annealed and then cooled. The process typically concludes with a series of cold rolling passes where each pass reduces the thickness of the plate by a substantially constant thickness amount to form the foil.

Post-irradiation examination of uranium 7 wt% molybdenum atomized dispersion fuel

NASA Astrophysics Data System (ADS)

Leenaers, A.; Van den Berghe, S.; Koonen, E.; Jarousse, C.; Huet, F.; Trotabas, M.; Boyard, M.; Guillot, S.; Sannen, L.; Verwerft, M.

2004-10-01

Two low-enriched uranium fuel plates consisting of U-7wt%Mo atomized powder dispersed in an aluminum matrix, have been irradiated in the FUTURE irradiation rig of the BR2 reactor at SCK•CEN. The plates were submitted to a heat flux of maximum 353 W/cm 2 while the surface cladding temperature is kept below 130 °C. After 40 full power days, visual examination and profilometry of the fuel plates revealed an increase of the plate thickness. In view of this observation, the irradiation campaign was prematurely stopped and the fuel plates were retrieved from the reactor, having at their end-of-life a maximum burn-up of 32.8% 235U (6.5% FIMA). The microstructure of one of the fuel plates has been characterized in an extensive post-irradiation campaign. The U(Mo) fuel particles have been found to interact with the Al matrix, resulting in an interaction layer which can be identified as (U,Mo)Al 3 and (U,Mo)Al 4. Based on the composition of the interaction layer it is shown that the observed physical parameters like thickness of the interaction layer between the Al matrix and the U(Mo) fuel particles compare well to the values calculated by the MAIA code, an U(Mo) behavior modeling code developed by the Commissariat à l'énergie atomique (CEA).

Irradiation testing of high density uranium alloy dispersion fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

1997-10-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U{sub 2}Mo, or U{sub 3}Si{sub 2}. These experiments will be discharged at peak fuel burnups ofmore » 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.« less

Microstructure of RERTR DU-Alloys Irradiated with Krypton Ions

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Gan; D. Keiser; D. Wachs

2009-11-01

Fuel development for reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium fuels that can be employed to replace existing high enrichment uranium fuels currently used in many research and test reactors worldwide. Radiation stability of the interaction product formed at fuel-matrix interface has a strong impact on fuel performance. Three depleted uranium alloys are cast that consist of the following 5 phases of interest to be investigated: U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, U6Mo4Al43 and UAl4. Irradiation of TEM disc samples with 500 keV Kr ions at 200?C to high doses up tomore » ~100 dpa were conducted using an intermediate voltage electron microscope equipped with an ion accelerator. The irradiated microstructure of the 5 phases is characterized using transmission electron microscopy. The results will be presented and the implication of the observed irradiated microstructure on the fuel performance will be discussed.« less

Investigation of welding and brazing of molybdenum and TZM alloy tubes

NASA Technical Reports Server (NTRS)

Lundblad, Wayne E.

1991-01-01

This effort involved investigating the welding and brazing techniques of molybdenum tubes to be used as cartridges in the crystal growth cartridge. Information is given in the form of charts and photomicrographs. It was found that the recrystallization temperature of molybdenum can be increased by alloying it with 0.5 percent titanium and 0.1 percent zirconium. Recrystallization temperatures for this alloy, known as TZM, become significant around 2500 F. A series of microhardness tests were run on samples of virgin and heat soaked TZM. The test results are given in tabular form. It was concluded that powder metallurgy TZM may be an acceptable cartridge material.

Molybdenum-UO2 cerment irradiation at 1145 K

NASA Technical Reports Server (NTRS)

Mcdonald, G.

1971-01-01

Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.

Effect of molybdenum ion implantation of the pitting corrosion of depleted uranium - 0.75 titanium alloy. (Reannouncement with new availability information). Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lei, K.S.; Chang, F.; Levy, M.

1993-07-01

Pitting corrosion of molybdenum-ion-implanted, depleted uranium -0 75 Ti (DU -0 75 Ti) has been studied electrochemically in acidic, neutral, and alkaline solutions containing sodium chloride, and the results have been compared to those of the unimplanted DU -0 75 Ti. The data show that Mo implantation shifts the pitting potential of DU -0 75 Ti in the noble direction in acidic and alkaline solutions. In neutral 50 ppm Cl- solution, however, there is no beneficial effect of Mo implantation. Auger analysis studies show that before exposure to the solutions, all the molybdenum is in the oxide, which is approximatelymore » l000 A thick. After electrochemical scans in the acidic and alkaline chloride solutions, most of the Mo disappears from the oxide. However, no decrease in Mo concentration is found after exposure in neutral chloride solution. It is proposed that the implanted molybdenum dissolves in the acidic and alkaline solutions and forms simple or complex molybdates that inhibit pitting corrosion. The implanted molybdenum does not dissolve in the neutral chloride solution and inhibition does not occur.« less

Electrochemical method of producing eutectic uranium alloy and apparatus

DOEpatents

Horton, James A.; Hayden, H. Wayne

1995-01-01

An apparatus and method for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride. The method of the invention produces an eutectic uranium alloy by creating an electrolyte reservoir defined by a container comprising an anode, placing an electrolyte in the reservoir, the electrolyte comprising a chlorine or fluorine containing salt and uranium chloride in molten form, positioning a cathode in the reservoir where the cathode comprises a metal capable of forming an uranium alloy having a melting point less than the melting point of pure uranium, and applying a current between the cathode and the anode.

Electrochemical method of producing eutectic uranium alloy and apparatus

DOEpatents

Horton, J.A.; Hayden, H.W.

1995-01-10

An apparatus and method are disclosed for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride. The method of the invention produces an eutectic uranium alloy by creating an electrolyte reservoir defined by a container comprising an anode, placing an electrolyte in the reservoir, the electrolyte comprising a chlorine or fluorine containing salt and uranium chloride in molten form, positioning a cathode in the reservoir where the cathode comprises a metal capable of forming an uranium alloy having a melting point less than the melting point of pure uranium, and applying a current between the cathode and the anode. 2 figures.

Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okuniewski, Maria A.; Ganapathy, Varsha; Hamilton, Brenden

2016-09-01

ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated tomore » 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.« less

Interdiffusion, Intrinsic Diffusion, Atomic Mobility, and Vacancy Wind Effect in γ(bcc) Uranium-Molybdenum Alloy

NASA Astrophysics Data System (ADS)

Huang, Ke; Keiser, Dennis D.; Sohn, Yongho

2013-02-01

U-Mo alloys are being developed as low enrichment uranium fuels under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. In order to understand the fundamental diffusion behavior of this system, solid-to-solid pure U vs Mo diffusion couples were assembled and annealed at 923 K, 973 K, 1073 K, 1173 K, and 1273 K (650 °C, 700 °C, 800 °C, 900 °C, and 1000 °C) for various times. The interdiffusion microstructures and concentration profiles were examined via scanning electron microscopy and electron probe microanalysis, respectively. As the Mo concentration increased from 2 to 26 at. pct, the interdiffusion coefficient decreased, while the activation energy increased. A Kirkendall marker plane was clearly identified in each diffusion couple and utilized to determine intrinsic diffusion coefficients. Uranium intrinsically diffused 5-10 times faster than Mo. Molar excess Gibbs free energy of U-Mo alloy was applied to calculate the thermodynamic factor using ideal, regular, and subregular solution models. Based on the intrinsic diffusion coefficients and thermodynamic factors, Manning's formalism was used to calculate the tracer diffusion coefficients, atomic mobilities, and vacancy wind parameters of U and Mo at the marker composition. The tracer diffusion coefficients and atomic mobilities of U were about five times larger than those of Mo, and the vacancy wind effect increased the intrinsic flux of U by approximately 30 pct.

ELECTRO-DEPOSITION OF NICKEL ALLOYS FROM THE PYROPHOSPHATE BATH: NICKEL- ZINC AND NICKEL-MOLYBDENUM ALLOYS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Panikkar, S.K.; Char, T.L.R.

1958-02-01

Results of studies on the electrodeposition of nickel-zinc and nickel-- molybdenum alloys in a pyrophosphate bath using platinium electrodes are presented. The fects of varying current density and metal contents of the electrolyte on alloy deposit composition, cathode efficiency, and cathode potential are presented in tabular form. (J.R.D.) l2432 A study was made of the effect of homogenization on the mechanical properties of solution-treated and aged aluminum and the quantitative effects of several variables on hardness. The effect of alloying elements on the increase in hardness of aluminum is shown. (J.E.D.)

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding andmore » Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.« less

Reactive melt infiltration of silicon-molybdenum alloys into microporous carbon preforms

NASA Technical Reports Server (NTRS)

Singh, M.; Behrendt, D. R.

1995-01-01

Investigations on the reactive melt infiltration of silicon-1.7 and 3.2 at.% molybdenum alloys into microporous carbon preforms have been carried out by modeling, differential thermal analysis (DTA), and melt infiltration experiments. These results indicate that the pore volume fraction of the carbon preform is a very important parameter in determining the final composition of the reaction-formed silicon carbide and the secondary phases. Various undesirable melt infiltration results, e.g. choking-off, specimen cracking, silicon veins, and lake formation, and their correlation with inadequate preform properties are presented. The liquid silicon-carbon reaction exotherm temperatures are influenced by the pore and carbon particle size of the preform and the compositions of infiltrants. Room temperature flexural strength and fracture toughness of materials made by the silicon-3.2 at.% molybdenum alloy infiltration of medium pore size preforms are also discussed.

Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

PubMed

Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

2002-12-01

This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

TERNARY ALLOYS OF URANIUM, COLUMBIUM, AND ZIRCONIUM

DOEpatents

Foote, F.G.

1960-08-01

Ternary alloys of uranium are described which are useful as neutron- reflecting materials in a fast neutron reactor. They are especially resistant to corrosion caused by oxidative processes of gascous or aqueous origin and comprise uranium as the predominant metal with zirconiunn and niobium wherein the total content of the minor alloying elements is between 2 and 8% by weight.

IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

M.K. Meyer; J. Gan; J.-F. Jue

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

First-principles studies of chromium line-ordered alloys in a molybdenum disulfide monolayer

NASA Astrophysics Data System (ADS)

Andriambelaza, N. F.; Mapasha, R. E.; Chetty, N.

2017-08-01

Density functional theory calculations have been performed to study the thermodynamic stability, structural and electronic properties of various chromium (Cr) line-ordered alloy configurations in a molybdenum disulfide (MoS2) hexagonal monolayer for band gap engineering. Only the molybdenum (Mo) sites were substituted at each concentration in this study. For comparison purposes, different Cr line-ordered alloy and random alloy configurations were studied and the most thermodynamically stable ones at each concentration were identified. The configurations formed by the nearest neighbor pair of Cr atoms are energetically most favorable. The line-ordered alloys are constantly lower in formation energy than the random alloys at each concentration. An increase in Cr concentration reduces the lattice constant of the MoS2 system following the Vegard’s law. From density of states analysis, we found that the MoS2 band gap is tunable by both the Cr line-ordered alloys and random alloys with the same magnitudes. The reduction of the band gap is mainly due to the hybridization of the Cr 3d and Mo 4d orbitals at the vicinity of the band edges. The band gap engineering and magnitudes (1.65 eV to 0.86 eV) suggest that the Cr alloys in a MoS2 monolayer are good candidates for nanotechnology devices.

Tensile and creep properties of titanium-vanadium, titanium-molybdenum, and titanium-niobium alloys

NASA Technical Reports Server (NTRS)

Gray, H. R.

1975-01-01

Tensile and creep properties of experimental beta-titanium alloys were determined. Titanium-vanadium alloys had substantially greater tensile and creep strength than the titanium-niobium and titanium-molybdenum alloys tested. Specific tensile strengths of several titanium-vanadium-aluminum-silicon alloys were equivalent or superior to those of commercial titanium alloys to temperatures of 650 C. The Ti-50V-3Al-1Si alloy had the best balance of tensile strength, creep strength, and metallurgical stability. Its 500 C creep strength was far superior to that of a widely used commercial titanium alloy, Ti-6Al-4V, and almost equivalent to that of newly developed commercial titanium alloys.

Molybdenum isotope fractionation during acid leaching of a granitic uranium ore

NASA Astrophysics Data System (ADS)

Migeon, Valérie; Bourdon, Bernard; Pili, Eric; Fitoussi, Caroline

2018-06-01

As an attempt to prevent illicit trafficking of nuclear materials, it is critical to identify the origin and transformation of uranium materials from the nuclear fuel cycle based on chemical and isotope tracers. The potential of molybdenum (Mo) isotopes as tracers is considered in this study. We focused on leaching, the first industrial process used to release uranium from ores, which is also known to extract Mo depending on chemical conditions. Batch experiments were performed in the laboratory with pH ranging from 0.3 to 5.5 in sulfuric acid. In order to span a large range in uranium and molybdenum yields, oxidizers such as nitric acid, hydrogen peroxide and manganese dioxide were also added. An enrichment in heavy Mo isotopes is produced in the solution during leaching of a granitic uranium ore, when Mo recovery is not quantitative. At least two Mo reservoirs were identified in the ore: ∼40% as Mo oxides soluble in water or sulfuric acid, and ∼40% of Mo hosted in sulfides soluble in nitric acid or hydrogen peroxide. At pH > 1.8, adsorption and/or precipitation processes induce a decrease in Mo yields with time correlated with large Mo isotope fractionations. Quantitative models were used to evaluate the relative importance of the processes involved in Mo isotope fractionation: dissolution, adsorption, desorption, precipitation, polymerization and depolymerization. Model best fits are obtained when combining the effects of dissolution/precipitation, and adsorption/desorption onto secondary minerals. These processes are inferred to produce an equilibrium isotope fractionation, with an enrichment in heavy Mo isotopes in the liquid phase and in light isotopes in the solid phase. Quantification of Mo isotope fractionation resulting from uranium leaching is thus a promising tool to trace the origin and transformation of nuclear materials. Our observations of Mo leaching are also consistent with observations of natural Mo isotope fractionation taking place during

FABRICATION OF URANIUM-ALUMINUM ALLOYS

DOEpatents

Saller, H.A.

1959-12-15

A process is presented for producing a workable article of a uranium- aluminum alloy in which the uranium content is between 14 and 70% by weight; aluminum powder and powdered UAl/sub 2/, UAl/sub 3/, UAl/sub 5/, or UBe/sub 9/ are mixed, and the mixture is compressed into the shape desired and sintered at between 450 and 600 deg C.

Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

NASA Astrophysics Data System (ADS)

Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

2014-09-01

One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo99 used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 106 cm-1) in a tube, their delta reactivities are the still

Procedure for Uranium-Molybdenum Density Measurements and Porosity Determination

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prabhakaran, Ramprashad; Devaraj, Arun; Joshi, Vineet V.

2016-08-13

The purpose of this document is to provide guidelines for preparing uranium-molybdenum (U-Mo) specimens, performing density measurements, and computing sample porosity. Typical specimens (solids) will be sheared to small rectangular foils, disks, or pieces of metal. A mass balance, solid density determination kit, and a liquid of known density will be used to determine the density of U-Mo specimens using the Archimedes principle. A standard test weight of known density would be used to verify proper operation of the system. By measuring the density of a U-Mo sample, it is possible to determine its porosity.

Irradiation performance of U-Mo monolithic fuel

DOE PAGES

Meyer, M. K.; Gan, J.; Jue, J. F.; ...

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

A K Wertsching

2012-09-01

As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Fluxmore » Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias

Simultaneous material flow analysis of nickel, chromium, and molybdenum used in alloy steel by means of input-output analysis.

PubMed

Nakajima, Kenichi; Ohno, Hajime; Kondo, Yasushi; Matsubae, Kazuyo; Takeda, Osamu; Miki, Takahiro; Nakamura, Shinichiro; Nagasaka, Tetsuya

2013-05-07

Steel is not elemental iron but rather a group of iron-based alloys containing many elements, especially chromium, nickel, and molybdenum. Steel recycling is expected to promote efficient resource use. However, open-loop recycling of steel could result in quality loss of nickel and molybdenum and/or material loss of chromium. Knowledge about alloying element substance flow is needed to avoid such losses. Material flow analyses (MFAs) indicate the importance of steel recycling to recovery of alloying elements. Flows of nickel, chromium, and molybdenum are interconnected, but MFAs have paid little attention to the interconnected flow of materials/substances in supply chains. This study combined a waste input-output material flow model and physical unit input-output analysis to perform a simultaneous MFA for nickel, chromium, and molybdenum in the Japanese economy in 2000. Results indicated the importance of recovery of these elements in recycling policies for end-of-life (EoL) vehicles and constructions. Improvement in EoL sorting technologies and implementation of designs for recycling/disassembly at the manufacturing phase are needed. Possible solutions include development of sorting processes for steel scrap and introduction of easier methods for identifying the composition of secondary resources. Recovery of steel scrap with a high alloy content will reduce primary inputs of alloying elements and contribute to more efficient resource use.

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

A novel route for processing cobalt–chromium–molybdenum orthopaedic alloys

PubMed Central

Patel, Bhairav; Inam, Fawad; Reece, Mike; Edirisinghe, Mohan; Bonfield, William; Huang, Jie; Angadji, Arash

2010-01-01

Spark plasma sintering has been used for the first time to prepare the ASTM F75 cobalt–chromium–molybdenum (Co–Cr–Mo) orthopaedic alloy composition using nanopowders. In the preliminary work presented in this report, the effect of processing variables on the structural features of the alloy (phases present, grain size and microstructure) has been investigated. Specimens of greater than 99.5 per cent theoretical density were obtained. Carbide phases were not detected in the microstructure but oxides were present. However, harder materials with finer grains were produced, compared with the commonly used cast/wrought processing methods, probably because of the presence of oxides in the microstructure. PMID:20200035

A simple spectrophotometric method for determination of zirconium or hafnium in selected molybdenum-base alloys

NASA Technical Reports Server (NTRS)

Dupraw, W. A.

1972-01-01

A simple analytical procedure is described for accurately and precisely determining the zirconium or hafnium content of molybdenum-base alloys. The procedure is based on the reaction of the reagent Arsenazo III with zirconium or hafnium in strong hydrochloric acid solution. The colored complexes of zirconium or hafnium are formed in the presence of molybdenum. Titanium or rhenium in the alloy have no adverse effect on the zirconium or hafnium complex at the following levels in the selected aliquot: Mo, 10 mg; Re, 10 mg; Ti, 1 mg. The spectrophotometric measurement of the zirconium or hafnium complex is accomplished without prior separation with a relative standard deviation of 1.3 to 2.7 percent.

Use of ion beams to simulate reaction of reactor fuels with their cladding

NASA Astrophysics Data System (ADS)

Birtcher, R. C.; Baldo, P.

2006-01-01

Processes occurring within reactor cores are not amenable to direct experimental observation. Among major concerns are damage, fission gas accumulation and reaction between the fuel and its cladding all of which lead to swelling. These questions can be investigated through simulation with ion beams. As an example, we discuss the irradiation driven interaction of uranium-molybdenum alloys, intended for use as low-enrichment reactor fuels, with aluminum, which is used as fuel cladding. Uranium-molybdenum coated with a 100 nm thin film of aluminum was irradiated with 3 MeV Kr ions to simulate fission fragment damage. Mixing and diffusion of aluminum was followed as a function of irradiation with RBS and nuclear reaction analysis using the 27Al(p,γ)28Si reaction which occurs at a proton energy of 991.9 keV. During irradiation at 150 °C, aluminum diffused into the uranium alloy at a irradiation driven diffusion rate of 30 nm2/dpa. At a dose of 90 dpa, uranium diffusion into the aluminum layer resulted in formation of an aluminide phase at the initial interface. The thickness of this phase grew until it consumed the aluminum layer. The rapid diffusion of Al into these reactor fuels may offer explanation of the observation that porosity is not observed in the fuel particles but on their periphery.

METHOD OF SEPARATING URANIUM FROM ALLOYS

DOEpatents

Chiotti, P.; Shoemaker, H.E.

1960-06-28

Uranium can be recovered from metallic uraniumthorium mixtures containing uranium in comparatively small amounts. The method of recovery comprises adding a quantity of magnesium to a mass to obtain a content of from 48 to 85% by weight; melting and forming a magnesium-thorium alloy at a temperature of between 585 and 800 deg C; agitating the mixture, allowing the mixture to settle whereby two phases, a thorium-containing magnesium-rich liquid phase and a solid uranium-rich phase, are formed; and separating the two phases.

Hydrothermal uranium deposits containing molybdenum and fluorite in the Marysvale volcanic field, west-central Utah

USGS Publications Warehouse

Cunningham, C.G.; Rasmussen, J.D.; Steven, T.A.; Rye, R.O.; Rowley, P.D.; Romberger, S.B.; Selverstone, J.

1998-01-01

Uranium deposits containing molybdenum and fluorite occur in the Central Mining Area, near Marysvale, Utah, and formed in an epithermal vein system that is part of a volcanic/hypabyssal complex. They represent a known, but uncommon, type of deposit; relative to other commonly described volcanic-related uranium deposits, they are young, well-exposed and well-documented. Hydrothermal uranium-bearing quartz and fluorite veins are exposed over a 300 m vertical range in the mines. Molybdenum, as jordisite (amorphous MoS2, together with fluorite and pyrite, increase with depth, and uranium decreases with depth. The veins cut 23-Ma quartz monzonite, 20-Ma granite, and 19-Ma rhyolite ash-flow tuff. The veins formed at 19-18 Ma in a 1 km2 area, above a cupola of a composite, recurrent, magma chamber at least 24 ?? 5 km across that fed a sequence of 21- to 14-Ma hypabyssal granitic stocks, rhyolite lava flows, ash-flow tuffs, and volcanic domes. Formation of the Central Mining Area began when the intrusion of a rhyolite stock, and related molybdenite-bearing, uranium-rich, glassy rhyolite dikes, lifted the fractured roof above the stock. A breccia pipe formed and relieved magmatic pressures, and as blocks of the fractured roof began to settle back in place, flat-lying, concave-downward, 'pull-apart' fractures were formed. Uranium-bearing, quartz and fluorite veins were deposited by a shallow hydrothermal system in the disarticulated carapace. The veins, which filled open spaces along the high-angle fault zones and flat-lying fractures, were deposited within 115 m of the ground surface above the concealed rhyolite stock. Hydrothermal fluids with temperatures near 200??C, ??18OH2O ~ -1.5, ?? -1.5, ??DH2O ~ -130, log fO2 about -47 to -50, and pH about 6 to 7, permeated the fractured rocks; these fluids were rich in fluorine, molybdenum, potassium, and hydrogen sulfide, and contained uranium as fluoride complexes. The hydrothermal fluids reacted with the wallrock resulting in

Biocompatibility and characterization of a Kolsterised(®) medical grade cobalt-chromium-molybdenum alloy.

PubMed

Conti, Malcolm Caligari; Karl, Andreas; Wismayer, Pierre Schembri; Buhagiar, Joseph

2014-01-01

High failure rates of cobalt-chromium-molybdenum (Co-Cr-Mo) metal-on-metal hip prosthesis were reported by various authors, probably due to the alloy's limited hardness and tribological properties. This thus caused the popularity of the alloy in metal-on-metal hip replacements to decrease due to its poor wear properties when compared with other systems such as ceramic-on-ceramic. S-phase surface engineering has become an industry standard when citing surface hardening of austenitic stainless steels. This hardening process allows the austenitic stainless steel to retain its corrosion resistance, while at the same time also improving its hardness and wear resistance. By coupling S-phase surface engineering, using the proprietary Kolsterising(®) treatment from Bodycote Hardiff GmbH, that is currently being used mainly on stainless steel, with Co-Cr-Mo alloys, an improvement in hardness and tribological characteristics is predicted. The objective of this paper is to analyze the biocompatibility of a Kolsterised(®) Co-Cr-Mo alloy, and to characterize the material surface in order to show the advantages gained by using the Kolsterised(®) material relative to the original untreated alloy, and other materials. This work has been performed on 3 fronts including; Material characterization, "In-vitro" corrosion testing, and Biological testing conforming to BS EN ISO 10993-18:2009 - Biological evaluation of medical devices. Using these techniques, the Kolsterised(®) cobalt-chromium-molybdenum alloys were found to have good biocompatibility and an augmented corrosion resistance when compared with the untreated alloy. The Kolsterised(®) samples also showed a 150% increase in surface hardness over the untreated material thus predicting better wear properties.

Biocompatibility and characterization of a Kolsterised® medical grade cobalt-chromium-molybdenum alloy

PubMed Central

Conti, Malcolm Caligari; Karl, Andreas; Wismayer, Pierre Schembri; Buhagiar, Joseph

2014-01-01

High failure rates of cobalt-chromium-molybdenum (Co-Cr-Mo) metal-on-metal hip prosthesis were reported by various authors, probably due to the alloy's limited hardness and tribological properties. This thus caused the popularity of the alloy in metal-on-metal hip replacements to decrease due to its poor wear properties when compared with other systems such as ceramic-on-ceramic. S-phase surface engineering has become an industry standard when citing surface hardening of austenitic stainless steels. This hardening process allows the austenitic stainless steel to retain its corrosion resistance, while at the same time also improving its hardness and wear resistance. By coupling S-phase surface engineering, using the proprietary Kolsterising® treatment from Bodycote Hardiff GmbH, that is currently being used mainly on stainless steel, with Co-Cr-Mo alloys, an improvement in hardness and tribological characteristics is predicted. The objective of this paper is to analyze the biocompatibility of a Kolsterised® Co-Cr-Mo alloy, and to characterize the material surface in order to show the advantages gained by using the Kolsterised® material relative to the original untreated alloy, and other materials. This work has been performed on 3 fronts including; Material characterization, “In-vitro” corrosion testing, and Biological testing conforming to BS EN ISO 10993–18:2009 - Biological evaluation of medical devices. Using these techniques, the Kolsterised® cobalt-chromium-molybdenum alloys were found to have good biocompatibility and an augmented corrosion resistance when compared with the untreated alloy. The Kolsterised® samples also showed a 150% increase in surface hardness over the untreated material thus predicting better wear properties. PMID:24451266

Real-time monitoring of plutonium content in uranium-plutonium alloys

DOEpatents

Li, Shelly Xiaowei; Westphal, Brian Robert; Herrmann, Steven Douglas

2015-09-01

A method and device for the real-time, in-situ monitoring of Plutonium content in U--Pu Alloys comprising providing a crucible. The crucible has an interior non-reactive to a metallic U--Pu alloy within said interior of said crucible. The U--Pu alloy comprises metallic uranium and plutonium. The U--Pu alloy is heated to a liquid in an inert or reducing atmosphere. The heated U--Pu alloy is then cooled to a solid in an inert or reducing atmosphere. As the U--Pu alloy is cooled, the temperature of the U--Pu alloy is monitored. A solidification temperature signature is determined from the monitored temperature of the U--Pu alloy during the step of cooling. The amount of Uranium and the amount of Plutonium in the U--Pu alloy is then determined from the determined solidification temperature signature.

Characterization of deformation mechanisms in zirconium alloys: effect of temperature and irradiation

NASA Astrophysics Data System (ADS)

Long, Fei

Zirconium alloys have been widely used in the CANDU (CANada Deuterium Uranium) reactor as core structural materials. Alloy such as Zircaloy-2 has been used for calandria tubes; fuel cladding; the pressure tube is manufactured from alloy Zr-2.5Nb. During in-reactor service, these alloys are exposed to a high flux of fast neutron at elevated temperatures. It is important to understand the effect of temperature and irradiation on the deformation mechanism of zirconium alloys. Aiming to provide experimental guidance for future modeling predictions on the properties of zirconium alloys this thesis describes the result of an investigation of the change of slip and twinning modes in Zircaloy-2 and Zr-2.5Nb as a function of temperature and irradiation. The aim is to provide scientific fundamentals and experimental evidences for future industry modeling in processing technique design, and in-reactor property change prediction of zirconium components. In situ neutron diffraction mechanical tests carried out on alloy Zircaloy-2 at three temperatures: 100¢ªC, 300¢ªC, and 500¢ªC, and described in Chapter 3. The evolution of the lattice strain of individual grain families in the loading and Poisson's directions during deformation, which probes the operation of slip and twinning modes at different stress levels, are described. By using the same type of in situ neutron diffraction technique, tests on Zr-2.5Nb pressure tube material samples, in either the fast-neutron irradiated or un-irradiated condition, are reported in Chapter 4. In Chapter 5, the measurement of dislocation density by means of line profile analysis of neutron diffraction patterns, as well as TEM observations of the dislocation microstructural evolution, is described. In Chapter 6 a hot-rolled Zr-2.5Nb with a larger grain size compared with the pressure tubing was used to study the development of dislocation microstructures with increasing plastic strain. In Chapter 7, in situ loading of heavy ion

Thermodynamic properties of uranium in liquid gallium, indium and their alloys

NASA Astrophysics Data System (ADS)

Volkovich, V. A.; Maltsev, D. S.; Yamshchikov, L. F.; Osipenko, A. G.

2015-09-01

Activity, activity coefficients and solubility of uranium was determined in gallium, indium and gallium-indium alloys containing 21.8 (eutectic), 40 and 70 wt.% In. Activity was measured at 573-1073 K employing the electromotive force method, and solubility between room temperature (or the alloy melting point) and 1073 K employing direct physical measurements. Activity coefficients were obtained from the difference of experimentally determined temperature dependencies of uranium activity and solubility. Intermetallic compounds formed in the respective alloys were characterized using X-ray diffraction. Partial and excess thermodynamic functions of uranium in the studied alloys were calculated. Liquidus lines in U-Ga and U-In phase diagrams from the side rich in gallium or indium are proposed.

RECOVERY OF URANIUM FROM PITCHBLENDE

DOEpatents

Ruehle, A.E.

1958-06-24

The decontamination of uranium from molybdenum is described. When acid solutions containing uranyl nitrate are contacted with ether for the purpose of extracting the uranium values, complex molybdenum compounds are coextracted with the uranium and also again back-extracted from the ether with the uranium. This invention provides a process for extracting uranium in which coextraction of molybdenum is avoided. It has been found that polyhydric alcohols form complexes with molybdenum which are preferentially water-soluble are taken up by the ether extractant to only a very minor degree. The preferred embodiment of the process uses mannitol, sorbitol or a mixture of the two as the complexing agent.

TEM analysis of irradiation-induced interaction layers in coated UMo/X/Al trilayer systems (X= Ti, Nb, Zr, and Mo)

NASA Astrophysics Data System (ADS)

Chiang, H.-Y.; Wiss, T.; Park, S.-H.; Dieste-Blanco, O.; Petry, W.

2018-02-01

Uranium-molybdenum (UMo) alloy powder embedded in an Al matrix is considered as a promising candidate for fuel conversion of research reactors. A modified system with a diffusion barrier X as coating, UMo/X/Al trilayer (X = Ti, Zr, Nb, and Mo), has been investigated to suppress interdiffusion between UMo and the Al matrix. The trilayer systems were tested by swift heavy ion irradiation, the thereby created interaction zone has been analyzed by scanning transmission electron microscopy (STEM) and energy-dispersive X-ray spectroscopy (EDX). Detailed structural characterization are presented and compared to earlier μ-XRD analysis.

Distillation of cadmium from uranium plutonium cadmium alloy

NASA Astrophysics Data System (ADS)

Kato, Tetsuya; Iizuka, Masatoshi; Inoue, Tadashi; Iwai, Takashi; Arai, Yasuo

2005-04-01

Uranium-plutonium alloy was prepared by distillation of cadmium from U-Pu-Cd ternary alloy. The initial ternary alloy contained 2.9 wt% U and 8.7 wt% Pu other than Cd, which were recovered by molten salt electrolysis with liquid Cd cathode. The distillation experiments were conducted in 10 g scale of the initial alloy using a small-scale distillation furnace equipped with an evaporator and a condenser in a vacuum vessel. After distillation at 1073 K, the weight of the residue was in good agreement with that of the loaded actinides, where the content of Cd decreased to less than 0.05 wt%. The uranium-plutonium alloy product was recovered without adhering to the yttria crucible. The cross section of the product was observed using electron probe micro-analyzer and it was found to consist of a dense material. Almost all of the evaporated Cd was recovered in the condenser and so enclosed well in the apparatus.

Theoretical Model for Volume Fraction of UC, 235U Enrichment, and Effective Density of Final U 10Mo Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Devaraj, Arun; Prabhakaran, Ramprashad; Joshi, Vineet V.

2016-04-12

The purpose of this document is to provide a theoretical framework for (1) estimating uranium carbide (UC) volume fraction in a final alloy of uranium with 10 weight percent molybdenum (U-10Mo) as a function of final alloy carbon concentration, and (2) estimating effective 235U enrichment in the U-10Mo matrix after accounting for loss of 235U in forming UC. This report will also serve as a theoretical baseline for effective density of as-cast low-enriched U-10Mo alloy. Therefore, this report will serve as the baseline for quality control of final alloy carbon content

Mineral resource of the month: molybdenum

USGS Publications Warehouse

Magyar, Michael J.

2004-01-01

Molybdenum is a metallic element that is most frequently used in alloy and stainless steels, which together represent the single largest market for molybdenum. Molybdenum has also proven invaluable in carbon steel, cast iron and superalloys. Its alloying versatility is unmatched because its addition enhances material performance under high-stress conditions in expanded temperature ranges and in highly corrosive environments. The metal is also used in catalysts, other chemicals, lubricants and many other applications.

A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

NASA Astrophysics Data System (ADS)

Van Kleeck, Melissa A.

The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

Surface alloying of aluminum with molybdenum by high-current pulsed electron beam

NASA Astrophysics Data System (ADS)

Xia, Han; Zhang, Conglin; Lv, Peng; Cai, Jie; Jin, Yunxue; Guan, Qingfeng

2018-02-01

The surface alloying of pre-coated molybdenum (Mo) film on aluminum (Al) substrate by high-current pulsed electron beam (HCPEB) was investigated. The microstructure and phase analysis were conducted by X-ray diffraction (XRD), scanning electron microscopy (SEM) and transmission electron microscopy (TEM). The results show that Mo particles were dissolved into Al matrix to form alloying layer, which was composed of Mo, Al and acicular or equiaxed Al5Mo phases after surface alloying. Meanwhile, various structure defects such as dislocation loops, high-density dislocations and dislocation walls were observed in the alloying surface. The corrosion resistance was tested by using potentiodynamic polarization curves and electrochemical impedance spectra (EIS). Electrochemical results indicate that all the alloying samples had better corrosion resistance in 3.5 wt% NaCl solution compared to initial sample. The excellent corrosion resistance is mainly attributed to the combined effect of the structure defects and the addition of Mo element to form a more stable passive film.

SOLVENT EXTRACTION PROCESS FOR URANIUM RECOVERY

DOEpatents

Clark, H.M.; Duffey, D.

1958-06-17

A process is described for extracting uranium from uranium ore, wherein the uranium is substantially free from molybdenum contamination. In a solvent extraction process for recovering uranium, uranium and molybdenum ions are extracted from the ore with ether under high acidity conditions. The ether phase is then stripped with water at a lower controiled acidity, resaturated with salting materials such as sodium nitrate, and reextracted with the separation of the molybdenum from the uranium without interference from other metals that have been previously extracted.

The irradiation behavior of atomized U-Mo alloy fuels at high temperature

NASA Astrophysics Data System (ADS)

Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

2001-04-01

Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

Carbon nanotube synthesis via the catalytic chemical vapor deposition of methane in the presence of iron, molybdenum, and iron-molybdenum alloy thin layer catalysts

NASA Astrophysics Data System (ADS)

Yahyazadeh, Arash; Khoshandam, Behnam

In this study, we documented the catalytic chemical vapor deposition synthesis of carbon nanotubes (CNTs) using ferrocene and molybdenum hexacarbonyl as catalyst nanoparticle precursors and methane as a nontoxic and economical carbon source for the first time. Field emission scanning electron microscopy, energy dispersive X-ray spectroscopy, wavelength dispersive X-ray spectrometry and transmission electron microscopy of the thin layer catalyst as a simple and cost effective catalyst preparation after methane decomposition reaction, along with Fourier transform infrared spectroscopy and Raman spectroscopy confirmed the growth of CNTs, from bimetallic nanoparticles, which are converted into iron-molybdenum alloy nanoparticles at 700 °C for pretreatment by hydrogen after chemical vapor deposition of thin layers. An investigation of the weight percentages of the chemical elements present in the CNTs synthesized from iron-molybdenum catalyst using quartz sheet substrate at 750 °C, confirmed a significant carbon yield of 75.4% which represents high catalyst activity. Additionally, multi-walled carbon nanotubes (∼16-55 nm in diameter and 1.2 μm in length) were observed in the iron-molybdenum alloy sample after methane decomposition reaction at 750 °C for 35 min. To show the role of iron and molybdenum coated on silicon substrate as two thin layer catalysts, samples were considered for CNTs growth (diameter ∼47-69 nm) at 800 °C and 830 °C, respectively. Moreover, the effect of hydrogen pretreatment was evaluated in terms of active metal coating properly. The best graphitic structure due to Raman spectroscopy outcomes (ID/IG ratio) was obtained for iron coated on a quartz sheet, which was estimated at 0.8505. Thermogravimetric analysis proved the thermal stability of the synthesized CNTs using iron thin-layer catalyst up to 350 °C.

Some observations on uranium carbide alloy/tungsten compatibility

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1972-01-01

Chemical compatibility between both pure and thoriated tungsten and uranium carbide alloys was studied at 1800 C for up to 3300 hours. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, dependent upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. The presence of a thermal gradient had no effect on the reactions observed nor did the presence of thoria in the tungsten clad.

Some observations on uranium carbide alloy/tungsten compatibility.

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1972-01-01

Results of chemical compatibility tests between both pure tungsten and thoriated tungsten run at 1800 C for up to 3300 hours with uranium carbide alloys. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, depending upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. Neither the presence of a thermal gradient nor the presence of thoria in the tungsten clad affect the reactions observed.

Process for recovering niobium from uranium-niobium alloys

DOEpatents

Wallace, S.A.; Creech, E.T.; Northcutt, W.G.

1982-09-27

Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and form a precipitate of niobium stannide, then separating the precipitate from the acid.

Structure Evolution and Distributions of Grain-Boundary Misorientainons in Submicrocrystalline Molybdenum Irradiated with a Pulsed Electron Beam

NASA Astrophysics Data System (ADS)

Stepanova, E. N.; Grabovetskaya, G. P.; Teresov, A. D.; Mishin, I. P.

2018-05-01

Using the methods of electron backscatter diffraction, electron microscopy and X-ray diffraction analysis, it is demonstrated that irradiation of the surface of a submicrocrystalline molybdenum specimen with a pulsed electron beam in a non-melt regime results in the formation of a gradient structure in its bulk. The irradiation temperature is shown to affect the density of defects, the value of stress, and the distributions of grain-boundary misorientations in the surface and bulk of the submicrocrystalline molybdenum specimens.

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth

Grain boundaries in metallic alloys often play a crucial role, not only in determining the mechanical properties or thermal stability of alloys, but also in dictating the phase transformation kinetics during thermomechanical processing. We demonstrate that locally stabilized structure and compositional segregation at grain boundaries—“grain boundary complexions”—in a complex multicomponent alloy can be modified to influence the kinetics of cellular transformation during subsequent thermomechanical processing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography analysis of a metallic nuclear fuel highly relevant to worldwide nuclear non-proliferation efforts —uranium-10 wt% molybdenum (U-10Mo) alloy, new evidence for the existence of grainmore » boundary complexion is provided. We then modified the concentration of impurities dissolved in Υ-UMo grain interiors and/or segregated to Υ-UMo grain boundaries by changing the homogenization treatment, and these effects were used used to retard the kinetics of cellular transformation during subsequent sub-eutectoid annealing in this U-10-Mo alloy during sub-eutectoid annealing. Thus, this work provided insights on tailoring the final microstructure of the U-10Mo alloy, which can potentially improve the irradiation performance of this important class of alloy fuels.« less

Liquid uranium alloy-helium fission reactor

DOEpatents

Minkov, Vladimir

1986-01-01

This invention teaches a nuclear fission reactor having a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200.degree.-1800.degree. C. range, and even higher to 2500.degree. C., limited only by the thermal effectiveness of the structural materials, increasing the efficiency of power generation from the normal 30-35% with 300.degree.-500.degree. C. upper limit temperature to 50-65%. Irradiation of the circulating liquid fuel, as contrasted to only localized irradiation of a solid fuel, provides improved fuel utilization.

Process for recovering niobium from uranium-niobium alloys

DOEpatents

Wallace, Steven A.; Creech, Edward T.; Northcutt, Walter G.

1983-01-01

Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and leave an insoluble residue of niobium stannide, then separating the niobium stannide from the acid.

Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coffey, Greg W.; Meinhardt, Kerry D.; Joshi, Vineet V.

2015-03-01

The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce amore » uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is

Irradiation creep of dispersion strengthened copper alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

1997-04-01

Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed ontomore » the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.« less

Mineral resource of the month: molybdenum

USGS Publications Warehouse

Polyak, Désire E.

2011-01-01

The article offers information about the mineral molybdenum. Sources includes byproduct or coproduct copper-molybdenum deposits in the Western Cordillera of North and South America. Among the uses of molybdenum are stainless steel applications, as an alloy material for manufacturing vessels and as lubricants, pigments or chemicals. Also noted is the role played by molybdenum in renewable energy technology.

Molybdenum-rhenium alloy based high-Q superconducting microwave resonators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Singh, Vibhor, E-mail: v.singh@tudelft.nl; Schneider, Ben H.; Bosman, Sal J.

2014-12-01

Superconducting microwave resonators (SMRs) with high quality factors have become an important technology in a wide range of applications. Molybdenum-Rhenium (MoRe) is a disordered superconducting alloy with a noble surface chemistry and a relatively high transition temperature. These properties make it attractive for SMR applications, but characterization of MoRe SMR has not yet been reported. Here, we present the fabrication and characterization of SMR fabricated with a MoRe 60–40 alloy. At low drive powers, we observe internal quality-factors as high as 700 000. Temperature and power dependence of the internal quality-factors suggest the presence of the two level systems from themore » dielectric substrate dominating the internal loss at low temperatures. We further test the compatibility of these resonators with high temperature processes, such as for carbon nanotube chemical vapor deposition growth, and their performance in the magnetic field, an important characterization for hybrid systems.« less

Austenitic alloy and reactor components made thereof

DOEpatents

Bates, John F.; Brager, Howard R.; Korenko, Michael K.

1986-01-01

An austenitic stainless steel alloy is disclosed, having excellent fast neutron irradiation swelling resistance and good post irradiation ductility, making it especially useful for liquid metal fast breeder reactor applications. The alloy contains: about 0.04 to 0.09 wt. % carbon; about 1.5 to 2.5 wt. % manganese; about 0.5 to 1.6 wt. % silicon; about 0.030 to 0.08 wt. % phosphorus; about 13.3 to 16.5 wt. % chromium; about 13.7 to 16.0 wt. % nickel; about 1.0 to 3.0 wt. % molybdenum; and about 0.10 to 0.35 wt. % titanium.

THE ANALYSIS OF URANIUM-ZIRCONIUM ALLOYS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Milner, G.W.C.; Skewies, A.F.

1953-03-01

A satisfactory procedure is described for the analysis of uranium-zirconium alloys containing up to 25% zirconium. It is based on the separation of the zirconium from the uranium by dissolving the cupferron complex of the former element into chloroform. After the evaporation of the solvent from the combined organic extracts, the residue is ignited to zirconium oxide. The latter is then re-dissolved and zirconium is separated from other elements co-extracted in the solvent extraction procedure by precipitation with mandelic acid. The zirconium mandelate is finally ignited to oxide at 960 deg C. The uranium is separated from the aqueous solutionmore » remaining from the cupferron extraction by precipitating with tannin at a pH of 8; the precipitate being removed by filtration and then ignited a t 800 deg C. The residue is dissolved in nitric acid and the uranium is finally determined by precipitating as ammonium diuranate and then igniting to U{sub 3}O{sub 8}. (auth)« less

Carcinogenicity and Immunotoxicity of Embedded Depleted Uranium and Heavy-Metal Tungsten Alloy in Rodents

DTIC Science & Technology

2006-10-01

Embedded Depleted Uranium and Heavy-Metal Tungsten Alloy in Rodents PRINCIPAL INVESTIGATOR: John F. Kalinich, Ph.D...Carcinogenicity and Immunotoxicity of Embedded Depleted Uranium and Heavy- Metal Tungsten Alloy in Rodents 5b. GRANT NUMBER DAMD17-01-1-0821 5c...ABSTRACT This study investigated the carcinogenic and immunotoxic potential of embedded fragments of depleted uranium (DU) and a heavy-metal tungsten

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rabin, S.A.; Lotts, A.L.; Hammond, J.P.

Uranium --molybdenum alloy rods containing from 10 to 15 wt% Mo and 1/16- in. in diameter were successfully fabricated by hot rotary swaging, followed by machining to remove the protective sheathing (Inconel with molybdenum barrier). Structurally strong rods with densities greater than 95% of theoretical were produced from both calciumreduced uranium mixed with hydrogen-reduced molybdenum and acid-cleaned, prealloyed shot when reduced in area about 55% at 1050 or 1100 deg C. Alloy homogeneity was good with prealloyed powders; however, traces of molybdenum -rich, gamma phase persisted in the elemental uranium -molybdenum material after swaging at 1100 deg C. Swagings embodyingmore » hydride uranium or oxide- contaminated prealloyed shot were unsatisfactory because of insufficient consolidation or poor interparticle bonding. (auth)« less

RERTR-12 Insertion 2 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez; G. S. Chang; D. M. Wachs

2012-09-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

Method for fabricating uranium alloy articles without shape memory effects

DOEpatents

Banker, John G.

1985-01-01

Uranium-rich niobium and niobium-zirconium alloys possess a characteristic known as shape memory effect wherein shaped articles of these alloys recover their original shape when heated. The present invention circumvents this memory behavior by forming the alloys into the desired configuration at elevated temperatures with "cold" matched dies and maintaining the shaped articles between the dies until the articles cool to ambient temperature.

Method for fabricating uranium alloy articles without shape memory effects

DOEpatents

Banker, J.G.

1980-05-21

Uranium-rich niobium and niobium-zirconium alloys possess a characteristic known as shape memory effect wherein shaped articles of these alloys recover their original shape when heated. The present invention circumvents this memory behavior by forming the alloys into the desired configuration at elevated temperatures with cold matched dies and maintaining the shaped articles between the dies until the articles cool to ambient temperature.

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDeavitt, Sean M

2011-04-29

Helmreich outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors

The irradiation hardening of Ni-Mo-Cr and Ni-W-Cr alloy under Xe26+ ion irradiation

NASA Astrophysics Data System (ADS)

Chen, Huaican; Hai, Yang; Liu, Renduo; Jiang, Li; Ye, Xiang-xi; Li, Jianjian; Xue, Wandong; Wang, Wanxia; Tang, Ming; Yan, Long; Yin, Wen; Zhou, Xingtai

2018-04-01

The irradiation hardening of Ni-Mo-Cr and Ni-W-Cr alloy was investigated. 7 MeV Xe26+ ion irradiation was performed at room temperature and 650 °C with peak damage dose from 0.05 to 10 dpa. With the increase of damage dose, the hardness of Ni-Mo-Cr and Ni-W-Cr alloy increases, and reaches saturation at damage dose ≥1 dpa. Moreover, the damage dose dependence of hardness in both alloys can be described by the Makin and Minter's equation, where the effective critical volume of obstacles can be used to represent irradiation hardening resistance of the alloys. Our results also show that Ni-W-Cr alloy has better irradiation hardening resistance than Ni-Mo-Cr alloy. This is ascribed to the fact that the W, instead of Mo in the alloy, can suppress the formation of defects under ion irradiation.

Effects of combined silicon and molybdenum alloying on the size and evolution of microalloy precipitates in HSLA steels containing niobium and titanium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pavlina, Erik J., E-mail: e.pavlina@deakin.edu.au; Van Tyne, C.J.; Speer, J.G.

2015-04-15

The effects of combined silicon and molybdenum alloying additions on microalloy precipitate formation in austenite after single- and double-step deformations below the austenite no-recrystallization temperature were examined in high-strength low-alloy (HSLA) steels microalloyed with titanium and niobium. The precipitation sequence in austenite was evaluated following an interrupted thermomechanical processing simulation using transmission electron microscopy. Large (~ 105 nm), cuboidal titanium-rich nitride precipitates showed no evolution in size during reheating and simulated thermomechanical processing. The average size and size distribution of these precipitates were also not affected by the combined silicon and molybdenum additions or by deformation. Relatively fine (< 20more » nm), irregular-shaped niobium-rich carbonitride precipitates formed in austenite during isothermal holding at 1173 K. Based upon analysis that incorporated precipitate growth and coarsening models, the combined silicon and molybdenum additions were considered to increase the diffusivity of niobium in austenite by over 30% and result in coarser precipitates at 1173 K compared to the lower alloyed steel. Deformation decreased the size of the niobium-rich carbonitride precipitates that formed in austenite. - Highlights: • We examine combined Si and Mo additions on microalloy precipitation in austenite. • Precipitate size tends to decrease with increasing deformation steps. • Combined Si and Mo alloying additions increase the diffusivity of Nb in austenite.« less

RERTR-13 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez; M. A. Lillo; G. S. Chang

2012-09-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-13 was designed to assess performance of different types of neutron absorbers that can be potentially used as burnable poisons in the low enriched uranium-molybdenum based dispersion and monolithic fuels.1 The following report summarizes the life of the RERTR-13 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

Proton irradiation studies on Al and Al5083 alloy

NASA Astrophysics Data System (ADS)

Bhattacharyya, P.; Gayathri, N.; Bhattacharya, M.; Gupta, A. Dutta; Sarkar, Apu; Dhar, S.; Mitra, M. K.; Mukherjee, P.

2017-10-01

The change in the microstructural parameters and microhardness values in 6.5 MeV proton irradiated pure Al and Al5083 alloy samples have been evaluated using different model based techniques of X-ray diffraction Line Profile Analysis (XRD) and microindendation techniques. The detailed line profile analysis of the XRD data showed that the domain size increases and saturates with irradiation dose both in the case of Al and Al5083 alloy. The corresponding microstrain values did not show any change with irradiation dose in the case of the pure Al but showed an increase at higher irradiation doses in the case of Al5083 alloy. The microindendation results showed that unirradiated Al5083 alloy has higher hardness value compared to that of unirradiated pure Al. The hardness increased marginally with irradiation dose in the case of Al5083, whereas for pure Al, there was no significant change with dose.

Fusion neutron irradiation of Ni-Si alloys at high temperature*1

NASA Astrophysics Data System (ADS)

Huang, J. S.; Guinan, M. W.; Hahn, P. A.

1988-07-01

Two Ni-4% Si alloys, with different cold work levels, have been irradiated with 14-MeV fusion neutrons at 623 K, and their Curie temperatures have been monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2-MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14-MeV fusion neutrons is only 6-7% of that for an identical alloy irradiated by 2-MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6-7% for the fusion neutron irradiated sample.

Alloy softening in binary molybdenum alloys

NASA Technical Reports Server (NTRS)

Stephens, J. R.; Witzke, W. R.

1972-01-01

An investigation was conducted to determine the effects of alloy additions of Hf, Ta, W, Re, Os, Ir, and Pt on the hardness of Mo. Special emphasis was placed on alloy softening in these binary Mo alloys. Results showed that alloy softening was produced by those elements having an excess of s+d electrons compared to Mo, while those elements having an equal number or fewer s+d electrons than Mo failed to produce alloy softening. Alloy softening and hardening can be correlated with the difference in number of s+d electrons of the solute element and Mo.

Creep behavior of uranium carbide-based alloys

NASA Technical Reports Server (NTRS)

Seltzer, M. S.; Wright, T. R.; Moak, D. P.

1975-01-01

The present work gives the results of experiments on the influence of zirconium carbide and tungsten on the creep properties of uranium carbide. The creep behavior of high-density UC samples follows the classical time-dependence pattern of (1) an instantaneous deformation, (2) a primary creep region, and (3) a period of steady-state creep. Creep rates for unalloyed UC-1.01 and UC-1.05 are several orders of magnitude greater than those measured for carbide alloys containing a Zr-C and/or W dispersoid. The difference in creep strength between alloyed and unalloyed materials varies with temperature and applied stress.

Microplastic deformation of polycrystalline iron and molybdenum subjected to high-current electron-beam irradiation

NASA Astrophysics Data System (ADS)

Dudarev, E. F.; Pochivalova, G. P.; Proskurovskii, D. I.; Rotshtein, V. P.; Markov, A. B.

1996-03-01

A technique for determination of residual stresses at various distances from the irradiated surface is proposed. It is established for iron and molybdenum that compressive stresses are set up under irradiation by low-energy high-current electron beams and that their values decrease sharply with increasing distance from the surface. The residual stresses are much smaller in absolute magnitude than those operating during irradiation. It is shown that the change in resistance to microplastic deformation on irradiation with low-energy high-current electron beams is governed not only by formation of a gradient dislocation substructure in the surface layer, but also by the residual stresses and the appearance of the Bauschinger effect.

Comparison of spring characteristics of titanium-molybdenum alloy and stainless steel

PubMed Central

Salehi, Anahita; Asatourian, Armen

2017-01-01

Background Titanium-molybdenum alloy (TMA) and stainless steel (SS) wires are commonly used in orthodontics as arch-wires for tooth movement. However, plastic deformation phenomenon in these arch-wires seems to be a major concern among orthodontists. This study aimed to compare the mechanical properties of TMA and SS wires with different dimensions. Material and Methods Seventy-two wire samples (36 TMA and 36 SS) of three different sizes (19×25, 17×25 and 16×22) were analyzed in vitro, with 12 samples in each group. Various mechanical properties of the wires, including spring-back, bending moment and stiffness were determined using a universal testing machine. Student’s t-test showed statistically significant differences in the mean values of all the groups. In addition, metallographic comparison of SS and TMA wires was conducted under an optical microscope. Results The degree of stiffness of 16×22-sized SS and TMA springs was found to be 12±2 and 5±0.4, respectively, while the bending moment was estimated to be 1927±352 (gm-mm) and 932±16 (gm-mm), respectively; the spring-back index was determined to be 0.61±0.2 and 0.4±.09, respectively (p<0.001). There were no statistically significant differences in spring-back index in larger dimensions of the wires. Conclusions Systematic analysis indicated that springs made of TMA were superior compared to those made of SS. Although both from economic and functionality viewpoints the use of TMA is suggested, further clinical investigations are recommended. Key words:Bending moment, optical microscope, spring-back, stainless steel, stiffness, titanium‒molybdenum alloy. PMID:28149469

Tensile Properties of Molybdenum and Tungsten from 2500 to 3700 F

NASA Technical Reports Server (NTRS)

Hall, Robert W.; Sikora, Paul F.

1959-01-01

Specimens of commercially pure sintered tungsten, arc-cast unalloyed molybdenum, and two arc-cast molybdenum-base alloys (one with 0.5 percent titanium, the other with 0.46 percent titanium and 0.07 percent zirconium) were fabricated from 1/2-inch-diameter rolled or swaged bars. All specimens were evaluated in short-time tensile tests in the as-received condition, and all except the molybdenum-titanium-zirconium alloy were tested after a 30-minute recrystallization anneal at 3800 F in a vacuum of approximately 0.1 micron. Results showed that the tungsten was considerably stronger than either the arc-cast unalloyed molybdenum or the molybdenum-base alloys over the 2500 to 3700 F temperature range. Recrystallization of swaged tungsten at 3800 F considerably reduced its tensile strength at 2500 F. However, above 3100 F, the as-swaged tungsten specimens recrystallized during testing, and had about the same strength as when recrystallized at 3800 F before evaluation. The ductility of molybdenum-base materials was very high at all test temperatures; the ductility of tungsten decreased sharply above about 3120 F.

Bioaccessibility of micron-sized powder particles of molybdenum metal, iron metal, molybdenum oxides and ferromolybdenum--Importance of surface oxides.

PubMed

Mörsdorf, Alexander; Odnevall Wallinder, Inger; Hedberg, Yolanda

2015-08-01

The European chemical framework REACH requires that hazards and risks posed by chemicals, including alloys and metals, that are manufactured, imported or used in different products (substances or articles) are identified and proven safe for humans and the environment. Metals and alloys need hence to be investigated on their extent of released metals (bioaccessibility) in biologically relevant environments. Read-across from available studies may be used for similar materials. This study investigates the release of molybdenum and iron from powder particles of molybdenum metal (Mo), a ferromolybdenum alloy (FeMo), an iron metal powder (Fe), MoO2, and MoO3 in different synthetic body fluids of pH ranging from 1.5 to 7.4 and of different composition. Spectroscopic tools and cyclic voltammetry have been employed to characterize surface oxides, microscopy, light scattering and nitrogen absorption for particle characterization, and atomic absorption spectroscopy to quantify released amounts of metals. The release of molybdenum from the Mo powder generally increased with pH and was influenced by the fluid composition. The mixed iron and molybdenum surface oxide of the FeMo powder acted as a barrier both at acidic and weakly alkaline conditions. These findings underline the importance of the surface oxide characteristics for the bioaccessibility of metal alloys. Copyright © 2015 The Authors. Published by Elsevier Inc. All rights reserved.

Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

DOE PAGES

Field, Kevin G.; Hu, Xunxiang; Littrell, Kenneth C.; ...

2015-07-14

The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition.more » Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. Furthermore, a structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.« less

Quantitative in vivo biocompatibility of new ultralow-nickel cobalt-chromium-molybdenum alloys.

PubMed

Sonofuchi, Kazuaki; Hagiwara, Yoshihiro; Koizumi, Yuichiro; Chiba, Akihiko; Kawano, Mitsuko; Nakayama, Masafumi; Ogasawara, Kouetsu; Yabe, Yutaka; Itoi, Eiji

2016-09-01

Nickel (Ni) eluted from metallic biomaterials is widely accepted as a major cause of allergies and inflammation. To improve the safety of cobalt-chromium-molybdenum (Co-Cr-Mo) alloy implants, new ultralow-Ni Co-Cr-Mo alloys with and without zirconium (Zr) have been developed, with Ni contents of less than 0.01%. In the present study, we investigated the biocompatibility of these new alloys in vivo by subcutaneously implanting pure Ni, conventional Co-Cr-Mo, ultralow-Ni Co-Cr-Mo, and ultralow-Ni Co-Cr-Mo with Zr wires into the dorsal sides of mice. After 3 and 7 days, tissues around the wire were excised, and inflammation; the expression of IL-1β, IL-6, and TNF-α; and Ni, Co, Cr, and Mo ion release were analyzed using histological analyses, qRT-PCR, and inductively coupled plasma mass spectrometry (ICP-MS), respectively. Significantly larger amounts of Ni eluted from pure Ni wires than from the other wires, and the degree of inflammation depended on the amount of eluted Ni. Although no significant differences in inflammatory reactions were identified among new alloys and conventional Co-Cr-Mo alloys in histological and qRT-PCR analyses, ICP-MS analysis revealed that Ni ion elution from ultralow-Ni Co-Cr-Mo alloys with and without Zr was significantly lower than from conventional Co-Cr-Mo alloys. Our study, suggests that the present ultralow-Ni Co-Cr-Mo alloys with and without Zr have greater safety and utility than conventional Co-Cr-Mo alloys. © 2016 Orthopaedic Research Society. Published by Wiley Periodicals, Inc. J Orthop Res 34:1505-1513, 2016. © 2016 Orthopaedic Research Society. Published by Wiley Periodicals, Inc.

Effect of rhenium on the structure and properties of the weld metal of a molybdenum alloy

NASA Technical Reports Server (NTRS)

Dyachenko, V. V.; Morozov, B. P.; Tylkina, M. A.; Savitskiy, Y. M.; Nikishanov, V. V.

1984-01-01

The structure and properties of welds made in molybdenum alloy VM-1 as a function of rhenium concentrations in the weld metal were studied. Rhenium was introduced into the weld using rhenium wire and tape or wires of Mo-47Re and Mo-52Re alloys. The properties of the weld metal were studied by means of metallographic techniques, electron microscopy, X-ray analysis, and autoradiography. The plasticity of the weld metal sharply was found to increase with increasing concentration of rhenium up to 50%. During welding, a decarburization process was observed which was more pronounced at higher concentrations of rhenium.

Irradiation creep of various ferritic alloys irradiated at {approximately}400{degrees}C in the PFR and FFTF reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

1997-04-01

Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400{degrees}C. Depending on the alloy starting condition, these steels develop amore » variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 x 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.« less

Cost Estimate for Molybdenum and Tantalum Refractory Metal Alloy Flow Circuit Concepts

NASA Technical Reports Server (NTRS)

Hickman, Robert R.; Martin, James J.; Schmidt, George R.; Godfroy, Thomas J.; Bryhan, A.J.

2010-01-01

The Early Flight Fission-Test Facilities (EFF-TF) team at NASA Marshall Space Flight Center (MSFC) has been tasked by the Naval Reactors Prime Contract Team (NRPCT) to provide a cost and delivery rough order of magnitude estimate for a refractory metal-based lithium (Li) flow circuit. The design is based on the stainless steel Li flow circuit that is currently being assembled for an NRPCT task underway at the EFF-TF. While geometrically the flow circuit is not representative of a final flight prototype, knowledge has been gained to quantify (time and cost) the materials, manufacturing, fabrication, assembly, and operations to produce a testable configuration. This Technical Memorandum (TM) also identifies the following key issues that need to be addressed by the fabrication process: Alloy selection and forming, cost and availability, welding, bending, machining, assembly, and instrumentation. Several candidate materials were identified by NRPCT including molybdenum (Mo) alloy (Mo-47.5 %Re), tantalum (Ta) alloys (T-111, ASTAR-811C), and niobium (Nb) alloy (Nb-1 %Zr). This TM is focused only on the Mo and Ta alloys, since they are of higher concern to the ongoing effort. The initial estimate to complete a Mo-47%Re system ready for testing is =$9,000k over a period of 30 mo. The initial estimate to complete a T-111 or ASTAR-811C system ready for testing is =$12,000k over a period of 36 mo.

Improvement of wear resistance of plasma-sprayed molybdenum blend coatings

NASA Astrophysics Data System (ADS)

Ahn, Jeehoon; Hwang, Byoungchul; Lee, Sunghak

2005-06-01

The wear resistance of plasma sprayed molybdenum blend coatings applicable to synchronizer rings or piston rings was investigated in this study. Four spray powders, one of which was pure molybdenum and the others blended powders of bronze and aluminum-silicon alloy powders mixed with molybdenum powders, were sprayed on a low-carbon steel substrate by atmospheric plasma spraying. Microstructural analysis of the coatings showed that the phases formed during spraying were relatively homogeneously distributed in the molybdenum matrix. The wear test results revealed that the wear rate of all the coatings increased with increasing wear load and that the blended coatings exhibited better wear resistance than the pure molybdenum coating, although the hardness was lower. In the pure molybdenum coatings, splats were readily fractured, or cracks were initiated between splats under high wear loads, thereby leading to the decrease in wear resistance. On the other hand, the molybdenum coating blended with bronze and aluminum-silicon alloy powders exhibited excellent wear resistance because hard phases such as CuAl2 and Cu9Al4 formed inside the coating.

Chemical state of fission products in irradiated uranium carbide fuel

NASA Astrophysics Data System (ADS)

Arai, Yasuo; Iwai, Takashi; Ohmichi, Toshihiko

1987-12-01

The chemical state of fission products in irradiated uranium carbide fuel has been estimated by equilibrium calculation using the SOLGASMIX-PV program. Solid state fission products are distributed to the fuel matrix, ternary compounds, carbides of fission products and intermetallic compounds among the condensed phases appearing in the irradiated uranium carbide fuel. The chemical forms are influenced by burnup as well as stoichiometry of the fuel. The results of the present study almost agree with the experimental ones reported for burnup simulated carbides.

Effects of neutron irradiation on deformation behavior of nickel-base fastener alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bajaj, R.; Mills, W.J.; Kammenzind, B.F.

1999-07-01

This paper presents the effects of neutron irradiation on the fracture behavior and deformation microstructure of high-strength nickel-base alloy fastener materials, Alloy X-750 and Alloy 625. Alloy X-750 in the HTH condition, and Alloy 625 in the direct aged condition were irradiated to a fluence of 2.4x10{sup 20} n/cm{sup 2} at 264 C in the Advanced Test Reactor. Deformation structures at low strains were examined. It was previously shown that Alloy X-750 undergoes hardening, a significant degradation in ductility and an increase in intergranular fracture. In contrast, Alloy 625 had shown softening with a concomitant increase in ductility and transgranularmore » failure after irradiation. The deformation microstructures of the two alloys were also different. Alloy X-750 deformed by a planar slip mechanism with fine microcracks forming at the intersections of slip bands with grain boundaries. Alloy 625 showed much more homogeneous deformation with fine, closely spaced slip bands and an absence of microcracks. The mechanism(s) of irradiation assisted stress corrosion cracking (IASCC) are discussed.« less

Image fusion of Secondary Ion Mass Spectrometry and Energy-dispersive X-Ray Spectroscopy data for the characterization of uranium-molybdenum fuel foils

NASA Astrophysics Data System (ADS)

Willingham, David; Naes, Benjamin E.; Tarolli, Jay G.; Schemer-Kohrn, Alan; Rhodes, Mark; Dahl, Michael; Guzman, Anthony; Burkes, Douglas E.

2018-01-01

Uranium-molybdenum (U-Mo) monolithic fuels represent one option for converting civilian research and test reactors operating with high enriched uranium (HEU) to low enriched uranium (LEU), effectively reducing the threat of nuclear proliferation world-wide. However, processes associated with fabrication of U-Mo monolithic fuels result in regions of elemental heterogeneity, observed as bands traversing the cross-section of representative samples. Isotopic variations (e.g., 235U and 238U) could also be introduced because of associated processing steps, particularly since HEU feedstock is melted with natural or depleted uranium diluent to produce LEU. This study demonstrates the utility of correlative analysis of Energy-Dispersive X-ray Spectroscopy (EDS) and Secondary Ion Mass Spectrometry (SIMS) with their image data streams using image fusion, resulting in a comprehensive microanalytical characterization toolbox. Elemental and isotopic measurements were made on a sample from the Advanced Test Reactor (ATR) Full-sized plate In-center flux trap Position (AFIP)-7 experiment and compared to previous optical and electron microscopy results. The image fusion results are characteristic of SIMS isotopic maps, but with the spatial resolution of EDS images and, therefore, can be used to increase the effective spatial resolution of the SIMS imaging results to better understand homogeneity or heterogeneity that persists because of processing selections. Visual inspection using the image fusion methodology indicated slight variations in the 235U/238U ratio and quantitative analysis using the image intensities across several FoVs revealed an average 235U atom percent value of 17.9 ± 2.4%, which was indicative of a non-uniform U isotopic distribution in the area sampled. Further development of this capability is useful for understanding the connections between the properties of LEU fuel alternatives and the ability to predict performance under irradiation.

PROCESSING OF NEUTRON-IRRADIATED URANIUM

DOEpatents

Hopkins, H.H. Jr.

1960-09-01

An improved "Purex" process for separating uranium, plutonium, and fission products from nitric acid solutions of neutron-irradiated uranium is offered. Uranium is first extracted into tributyl phosphate (TBP) away from plutonium and fission products after adjustment of the acidity from 0.3 to 0.5 M and heating from 60 to 70 deg C. Coextracted plutonium, ruthenium, and fission products are fractionally removed from the TBP by three scrubbing steps with a 0.5 M nitric acid solution of ferrous sulfamate (FSA), from 3.5 to 5 M nitric acid, and water, respectively, and the purified uranium is finally recovered from the TBP by precipitation with an aqueous solution of oxalic acid. The plutonium in the 0.3 to 0.5 M acid solution is oxidized to the tetravalent state with sodium nitrite and extracted into TBP containing a small amount of dibutyl phosphate (DBP). Plutonium is then back-extracted from the TBP-DBP mixture with a nitric acid solution of FSA, reoxidized with sodium nitrite in the aqueous strip solution obtained, and once more extracted with TBP alone. Finally the plutonium is stripped from the TBP with dilute acid, and a portion of the strip solution thus obtained is recycled into the TBPDBP for further purification.

Effects of long-term aging on ductility of the columbium alloys C-103, Cb-1Zr, and Cb-752 and the molybdenum alloy Mo-TZM

NASA Technical Reports Server (NTRS)

Stephens, J. R.

1975-01-01

A program was conducted to determine if aging embrittlement occurs in the columbium alloys C-103, CB-1Zr, and Cb-752 or in the molybdenum alloy Mo-TZM. Results showed that aging embrittlement does not occur in C-103, Cb-1Zr, or Mo-TZM during long-term (1000 hr) aging at temperatures in the range 700 to 1025 C. In contrast, aging embrittlement did occur in the Cb-752 alloy after similar aging at 900 C. A critical combination of the solute additions W and Zr in Cb-752 led to Zr segregation at grain boundaries during long-term aging. This segregation subsequently resulted in embrittlement as indicated by an increase in the ductile-brittle transition temperature from below -1960 C to about -150 C.

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

NASA Technical Reports Server (NTRS)

Bowles, K. J.; Gluyas, R. E.

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

2016-05-01

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

A model to predict thermal conductivity of irradiated U–Mo dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

The Office of Materials Management and Minimization Reactor Conversion Program continues to develop existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. The program is focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layermore » formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.« less

Mechanical properties of neutron-irradiated model and commercial FeCrAl alloys

NASA Astrophysics Data System (ADS)

Field, Kevin G.; Briggs, Samuel A.; Sridharan, Kumar; Howard, Richard H.; Yamamoto, Yukinori

2017-06-01

The development and understanding of the mechanical properties of neutron-irradiated FeCrAl alloys is increasingly a critical need as these alloys continue to become more mature for nuclear reactor applications. This study focuses on the mechanical properties of model FeCrAl alloys and of a commercial FeCrAl alloy neutron-irradiated to up to 13.8 displacements per atom (dpa) at irradiation temperatures between 320 and 382 °C. Tensile tests were completed at room temperature and at 320 °C, and a subset of fractured tensile specimens was examined by scanning electron microscopy. Results showed typical radiation hardening and embrittlement indicative of high chromium ferritic alloys with strong chromium composition dependencies at lower doses. At and above 7.0 dpa, the mechanical properties saturated for both the commercial and model FeCrAl alloys, although brittle cleavage fracture was observed at the highest dose in the model FeCrAl alloy with the highest chromium content (18 wt %). The results suggest the composition and microstructure of FeCrAl alloys plays a critical role in the mechanical response of FeCrAl alloys irradiated near temperatures relevant to light water reactors.

Temperature dependent surface modification of molybdenum due to low energy He+ ion irradiation

NASA Astrophysics Data System (ADS)

Tripathi, J. K.; Novakowski, T. J.; Joseph, G.; Linke, J.; Hassanein, A.

2015-09-01

In this paper, we report on the temperature dependent surface modifications in molybdenum (Mo) samples due to 100 eV He+ ion irradiation in extreme conditions as a potential candidate to plasma-facing components in fusion devices alternative to tungsten. The Mo samples were irradiated at normal incidence, using an ion fluence of 2.6 × 1024 ions m-2 (with a flux of 7.2 × 1020 ions m-2 s-1). Surface modifications have been studied using high-resolution field emission scanning electron-(SEM) and atomic force (AFM) microscopy. At 773 K target temperature homogeneous evolution of molybdenum nanograins on the entire Mo surface were observed. However, at 823 K target temperature appearance of nano-pores and pin-holes nearby the grain boundaries, and Mo fuzz in patches were observed. The fuzz density increases significantly with target temperatures and continued until 973 K. However, at target temperatures beyond 973 K, counterintuitively, a sequential reduction in the fuzz density has been seen till 1073 K temperatures. At 1173 K and above temperatures, only molybdenum nano structures were observed. Our temperature dependent studies confirm a clear temperature widow, 823-1073 K, for Mo fuzz formation. Ex-situ high resolution X-ray photoelectron spectroscopy studies on Mo fuzzy samples show the evidence of MoO3 3d doublets. This elucidates that almost all the Mo fuzz were oxidized during open air exposure and are thick enough as well. Likewise the microscopy studies, the optical reflectivity measurements also show a sequential reduction in the reflectivity values (i.e., enhancement in the fuzz density) up to 973 K and after then a sequential enhancement in the reflectivity values (i.e., reduction in the fuzz density) with target temperatures. This is in well agreement with microscopy studies where we observed clear temperature window for Mo fuzz growth.

Compositional redistribution in alloy films under high-voltage electron microscope irradiation

NASA Astrophysics Data System (ADS)

Lam, Nghi Q.; Leaf, O. K.; Minkoff, M.

1983-10-01

The problem of nonequilibrium segregation in alloy films under high-voltage electron microscope (HVEM) irradiation at elevated temperatures is re-examined in the present work, taking into account the damage-rate gradients caused by radial variation in the electron flux. Axial and radial compositional redistributions in model solid solutions, representative of concentrated Ni-Cu, Ni-Al and Ni-Si alloys, were calculated as a function of time, temperature, and film thickness, using a kinetic theory of segregation in binary alloys. The numerical results were achieved by means of a new software package (DISPL2) for solving convection-diffusion-kinetics problems with general orthogonal geometries. It was found that HVEM irradiation-induced segregation in thin films consists of two stages. Initially, due to the proximity of the film surfaces as sinks for point defects, the usual axial segregation (to surfaces) occurs at relatively short irradiation times, and rapidly attains quasi-steady state. Then, radial segregation becomes more and more competitive, gradually affecting the kinetics of axial segregation. At a given temperature, the buildup time to steady state is much longer in the present situation than in the simple case of one-dimensional segregation with uniform defect production. Changes in the alloy composition occur in a much larger zone than the irradiated volume. As a result, the average alloy composition within the irradiated region can differ greatly from that of the unirradiated alloy. The present calculations may be useful in the interpretation of the kinetics of certain HVEM irradiation-induced processes in alloys.

Low activation ferritic alloys

DOEpatents

Gelles, David S.; Ghoniem, Nasr M.; Powell, Roger W.

1986-01-01

Low activation ferritic alloys, specifically bainitic and martensitic stainless steels, are described for use in the production of structural components for nuclear fusion reactors. They are designed specifically to achieve low activation characteristics suitable for efficient waste disposal. The alloys essentially exclude molybdenum, nickel, nitrogen and niobium. Strength is achieved by substituting vanadium, tungsten, and/or tantalum in place of the usual molybdenum content in such alloys.

Low activation ferritic alloys

DOEpatents

Gelles, D.S.; Ghoniem, N.M.; Powell, R.W.

1985-02-07

Low activation ferritic alloys, specifically bainitic and martensitic stainless steels, are described for use in the production of structural components for nuclear fusion reactors. They are designed specifically to achieve low activation characteristics suitable for efficient waste disposal. The alloys essentially exclude molybdenum, nickel, nitrogen and niobium. Strength is achieved by substituting vanadium, tungsten, and/or tantalum in place of the usual molybdenum content in such alloys.

Nanoparticles alloying in liquids: Laser-ablation-generated Ag or Pd nanoparticles and laser irradiation-induced AgPd nanoparticle alloying

NASA Astrophysics Data System (ADS)

Semaltianos, N. G.; Chassagnon, R.; Moutarlier, V.; Blondeau-Patissier, V.; Assoul, M.; Monteil, G.

2017-04-01

Laser irradiation of a mixture of single-element micro/nanomaterials may lead to their alloying and fabrication of multi-element structures. In addition to the laser induced alloying of particulates in the form of micro/nanopowders in ambient atmosphere (which forms the basis of the field of additive manufacturing technology), another interesting problem is the laser-induced alloying of a mixture of single-element nanoparticles in liquids since this process may lead to the direct fabrication of alloyed-nanoparticle colloidal solutions. In this work, bare-surface ligand-free Ag and Pd nanoparticles in solution were prepared by laser ablation of the corresponding bulk target materials, separately in water. The two solutions were mixed and the mixed solution was laser irradiated for different time durations in order to investigate the laser-induced nanoparticles alloying in liquid. Nanoparticles alloying and the formation of AgPd alloyed nanoparticles takes place with a decrease of the intensity of the surface-plasmon resonance peak of the Ag nanoparticles (at ∼405 nm) with the irradiation time while the low wavelength interband absorption peaks of either Ag or Pd nanoparticles remain unaffected by the irradiation for a time duration even as long as 30 min. The nanoalloys have lattice constants with values between those of the pure metals, which indicates that they consist of Ag and Pd in an approximately 1:1 ratio similar to the atomic composition of the starting mixed-nanoparticle solution. Formation of nanoparticle networks consisting of bimetallic alloyed nanoparticles and nanoparticles that remain as single elements (even after the end of the irradiation), joining together, are also formed. The binding energies of the 3d core electrons of both Ag and Pd nanoparticles shift to lower energies with the irradiation time, which is also a typical characteristic of AgPd alloyed nanoparticles. The mechanisms of nanoparticles alloying and network formation are also

Nanoparticles alloying in liquids: Laser-ablation-generated Ag or Pd nanoparticles and laser irradiation-induced AgPd nanoparticle alloying.

PubMed

Semaltianos, N G; Chassagnon, R; Moutarlier, V; Blondeau-Patissier, V; Assoul, M; Monteil, G

2017-04-18

Laser irradiation of a mixture of single-element micro/nanomaterials may lead to their alloying and fabrication of multi-element structures. In addition to the laser induced alloying of particulates in the form of micro/nanopowders in ambient atmosphere (which forms the basis of the field of additive manufacturing technology), another interesting problem is the laser-induced alloying of a mixture of single-element nanoparticles in liquids since this process may lead to the direct fabrication of alloyed-nanoparticle colloidal solutions. In this work, bare-surface ligand-free Ag and Pd nanoparticles in solution were prepared by laser ablation of the corresponding bulk target materials, separately in water. The two solutions were mixed and the mixed solution was laser irradiated for different time durations in order to investigate the laser-induced nanoparticles alloying in liquid. Nanoparticles alloying and the formation of AgPd alloyed nanoparticles takes place with a decrease of the intensity of the surface-plasmon resonance peak of the Ag nanoparticles (at ∼405 nm) with the irradiation time while the low wavelength interband absorption peaks of either Ag or Pd nanoparticles remain unaffected by the irradiation for a time duration even as long as 30 min. The nanoalloys have lattice constants with values between those of the pure metals, which indicates that they consist of Ag and Pd in an approximately 1:1 ratio similar to the atomic composition of the starting mixed-nanoparticle solution. Formation of nanoparticle networks consisting of bimetallic alloyed nanoparticles and nanoparticles that remain as single elements (even after the end of the irradiation), joining together, are also formed. The binding energies of the 3d core electrons of both Ag and Pd nanoparticles shift to lower energies with the irradiation time, which is also a typical characteristic of AgPd alloyed nanoparticles. The mechanisms of nanoparticles alloying and network formation are also

Tensile and stress-rupture behavior of hafnium carbide dispersed molybdenum and tungsten base alloy wires

NASA Technical Reports Server (NTRS)

Yun, Hee Mann; Titran, Robert H.

1993-01-01

The tensile strain rate sensitivity and the stress-rupture strength of Mo-base and W-base alloy wires, 380 microns in diameter, were determined over the temperature range from 1200 K to 1600 K. Three molybdenum alloy wires; Mo + 1.1w/o hafnium carbide (MoHfC), Mo + 25w/o W + 1.1w/o hafnium carbide (MoHfC+25W) and Mo + 45w/o W + 1.1w/o hafnium carbide (MoHfC+45W), and a W + 0.4w/o hafnium carbide (WHfC) tungsten alloy wire were evaluated. The tensile strength of all wires studied was found to have a positive strain rate sensitivity. The strain rate dependency increased with increasing temperature and is associated with grain broadening of the initial fibrous structures. The hafnium carbide dispersed W-base and Mo-base alloys have superior tensile and stress-rupture properties than those without HfC. On a density compensated basis the MoHfC wires exhibit superior tensile and stress-rupture strengths to the WHfC wires up to approximately 1400 K. Addition of tungsten in the Mo-alloy wires was found to increase the long-term stress rupture strength at temperatures above 1400 K. Theoretical calculations indicate that the strength and ductility advantage of the HfC dispersed alloy wires is due to the resistance to recrystallization imparted by the dispersoid.

Damage structures in fission-neutron irradiated Ni-based alloys at high temperatures

NASA Astrophysics Data System (ADS)

Yamakawa, K.; Shimomura, Y.

1999-01-01

The defects formed in Ni based (Ni-Si, Ni-Cu and Ni-Fe) alloys which were irradiated with fission-neutrons were examined by electron microscopy. Irradiations were carried out at 473 K and 573 K. In the 473 K irradiated specimens, a high density of large interstitial loops and small vacancy clusters with stacking fault tetrahedra (SFT) were observed. The number densities of these two types of defects did not strongly depend on the amount of solute atoms in each alloy. The density of the loops in Ni-Si alloys was much higher than those in Ni-Cu and Ni-Fe alloys, while the density of SFT only slightly depended on the kind of solute. Also, the size of the loops depended on the kinds and amounts of solute. In 573 K irradiated Ni-Cu specimens, a high density of dislocation lines developed during the growth of interstitial loops. In Ni-Si alloys, the number density and size of the interstitial loops changed as a function of the amount of solute. Voids were formed in Ni-Cu alloys but scarcely formed in Ni-Si alloys. The number density of voids was one hundredth of that of SFT observed in 473 K irradiated Ni-Cu alloys. Possible formation processes of interstitial loops, SFT dislocation lines and voids are discussed.

Modeling Early-Stage Processes of U-10 Wt.%Mo Alloy Using Integrated Computational Materials Engineering Concepts

NASA Astrophysics Data System (ADS)

Wang, Xiaowo; Xu, Zhijie; Soulami, Ayoub; Hu, Xiaohua; Lavender, Curt; Joshi, Vineet

2017-12-01

Low-enriched uranium alloyed with 10 wt.% molybdenum (U-10Mo) has been identified as a promising alternative to high-enriched uranium. Manufacturing U-10Mo alloy involves multiple complex thermomechanical processes that pose challenges for computational modeling. This paper describes the application of integrated computational materials engineering (ICME) concepts to integrate three individual modeling components, viz. homogenization, microstructure-based finite element method for hot rolling, and carbide particle distribution, to simulate the early-stage processes of U-10Mo alloy manufacture. The resulting integrated model enables information to be passed between different model components and leads to improved understanding of the evolution of the microstructure. This ICME approach is then used to predict the variation in the thickness of the Zircaloy-2 barrier as a function of the degree of homogenization and to analyze the carbide distribution, which can affect the recrystallization, hardness, and fracture properties of U-10Mo in subsequent processes.

Effects of self-irradiation in plutonium alloys

DOE PAGES

Chung, B. W.; Lema, K. E.; Allen, P. G.

2015-09-16

In this paper, we present updated results of self-irradiation effects on 238Pu-enriched 239Pu alloys measured by immersion density, dilatometry, and tensile tests. We obtained the self-irradiation equivalent time of nearly 200 years, nearly 100 years longer than in our previous papers. At this extended aging, we find the rate of decrease in density has slowed significantly, stabilizing around 15.73 g/cc, without signs of void swelling. The volume expansion measured at 35°C also shows apparent saturation at less than 0.25%. Quasi-static tensile measurement still show gradual increase in the strength of plutonium alloys with age.

Irradiation and Thermal Annealing Effects in Amorphous Magnetic Alloys.

NASA Astrophysics Data System (ADS)

Fisher, David G.

Irradiation with protons, electrons, and alpha particles produces effects in amorphous magnetic alloys (Fe(,x)Ni(,80)P(,20-y)B(,y), where x was 20, 27, 34, or 40 and y was either 6 or 20) that appear analogous to effects produced by thermal annealing. The work presented in this dissertation represents an extension of work performed by Franz('(1)) and/or Donnelly.('(2)) The work of Franz, Donnelly, and this author has been a coordinated investigation into various aspects of radiation damage and thermal annealing effects in the above-mentioned amorphous alloys' magnetic properties. Upon either irradiation or thermal annealing, the Curie temperature, T(,c), is enhanced in these alloys. Also the relative permeability, (mu)(,r), is raised as much as seven-fold. Electrolytic layer removal experiments on proton-irradiated (0.25-MeV) samples conclusively demonstrate that the particle irradiation does not merely heat the sample bulk. Annealing studies performed on both irradiated and as-quenched samples suggested, via T(,c) measurement, that a structural relaxation process had taken place. The structural relaxation takes place as a result of a macroscopic heating in the case of the annealed samples and it is postulated that the structural relaxation takes place as a result of a miroscopic heating about the particle track (thermal spike mechanism) in the case of the irradiated samples. This work also presents preliminary results concerning the influence of irradiation and thermal annealing on the crystallization process in these alloys. The results of DSC and electrical resistivity (above room temperature) are presented. Using electrical resistivity as an indicator, a series of isothermal recrystallization measurements were performed using samples of 2.25-MeV proton-irradiated, 200(DEGREES)C-annealed, and as-quenched Fe(,20)Ni(,60)P(,14)B(,6). The activation energy for the onset of recrystallization is 2.0 eV for as-quenched samples and is 5.3 eV for the irradiated and

Mechanical properties of neutron-irradiated model and commercial FeCrAl alloys

DOE PAGES

Field, Kevin G.; Briggs, Samuel A.; Sridharan, Kumar; ...

2017-03-28

The development and understanding of the mechanical properties of neutron-irradiated FeCrAl alloys is increasingly a critical need as these alloys continue to become more mature for nuclear reactor applications. This study focuses on the mechanical properties of model FeCrAl alloys and of a commercial FeCrAl alloy neutron-irradiated to up to 13.8 displacements per atom (dpa) at irradiation temperatures between 320 and 382 °C. Tensile tests were completed at room temperature and at 320 °C, and a subset of fractured tensile specimens was examined by scanning electron microscopy. Results showed typical radiation hardening and embrittlement indicative of high chromium ferritic alloysmore » with strong chromium composition dependencies at lower doses. At and above 7.0 dpa, the mechanical properties saturated for both the commercial and model FeCrAl alloys, although brittle cleavage fracture was observed at the highest dose in the model FeCrAl alloy with the highest chromium content (18 wt %). Finally, the results suggest the composition and microstructure of FeCrAl alloys plays a critical role in the mechanical response of FeCrAl alloys irradiated near temperatures relevant to light water reactors.« less

Irradiation of three T-111 clad uranium nitride fuel pins for 8070 hours at 990 C (1815 F)

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.; Gedeon, L.; Galbo, R. J.

1973-01-01

The design and successful operation of three tantalum alloy (Ta-8W-2Hf) clad uranium mononitride (UN) fuel pins irradiated for 8070 hr at 990 C (1815 F) is described. Two pin diameters having measured burnups of 0.47 and 0.90 uranium atom percent were tested. No clad failures or swelling was detected; however, postirradiation clad samples tested failed with 1 percent strain. The fuel density decrease was 2 percent, and the fission gas release was less than 0.05 percent. Isotropic fuel swelling, which averaged about 0.5 percent, was less than fuel pin assembly clearances. Thus the clad was not strained. Thermocouples with a modified hot zone operated at average temperatures to 1100 C (2012 F) without failure. Factors that influence the ability to maintain uniform clad temperature as well as the results of the heat transfer calculations are discussed.

Plasma Processing Systems for the Manufacture of Refractory Metals and their Alloys for Military Needs

DTIC Science & Technology

1978-10-09

melting point is around 4000*K. An exceedingly interesting feature of these solidification composites is the formation of fibrous MC type carbide ...the matrix could be refractory metal binary alloys with copper or uranium and the eutectic phase could be carbide of tungsten, * molybdenum, tantalum or...42 Accs -n or - *DTTI Tf Avn ! -7ll ’ i CrDi t , l’’*i,;. LIST OF FIGURES FIG. 1 Flow Diagram of Cemented Carbide Manufacture

Cytotoxicity of titanium and titanium alloying elements.

PubMed

Li, Y; Wong, C; Xiong, J; Hodgson, P; Wen, C

2010-05-01

It is commonly accepted that titanium and the titanium alloying elements of tantalum, niobium, zirconium, molybdenum, tin, and silicon are biocompatible. However, our research in the development of new titanium alloys for biomedical applications indicated that some titanium alloys containing molybdenum, niobium, and silicon produced by powder metallurgy show a certain degree of cytotoxicity. We hypothesized that the cytotoxicity is linked to the ion release from the metals. To prove this hypothesis, we assessed the cytotoxicity of titanium and titanium alloying elements in both forms of powder and bulk, using osteoblast-like SaOS(2) cells. Results indicated that the metal powders of titanium, niobium, molybdenum, and silicon are cytotoxic, and the bulk metals of silicon and molybdenum also showed cytotoxicity. Meanwhile, we established that the safe ion concentrations (below which the ion concentration is non-toxic) are 8.5, 15.5, 172.0, and 37,000.0 microg/L for molybdenum, titanium, niobium, and silicon, respectively.

Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature

NASA Technical Reports Server (NTRS)

Rohal, R. G.; Tambling, T. N.; Smith, R. L.

1973-01-01

Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.

Effects of alloying elements on the formation of < c >-component loops in Zr alloy Excel under heavy ion irradiation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Idrees, Yasir; Francis, Elisabeth M.; Yao, Zhongwen

2015-05-14

We report here the microstructural changes occurring in the zirconium alloy Excel (Zr-3.5 wt% Sn-0.8Nb-0.8Mo-0.2Fe) during heavy ion irradiation. In situ irradiation experiments were conducted at reactor operating temperatures on two Zr Excel alloy microstructures with different states of alloying elements, with the states achieved by different solution heat treatments. In the first case, the alloying elements were mostly concentrated in the beta (beta) phase, whereas, in the second case, large Zr-3(Mo,Nb,Fe)(4) secondary phase precipitates (SPPs) were grown in the alpha (alpha) phase by long term aging. The heavy ion induced damage and resultant compositional changes were examined using transmissionmore » electron microscopy (TEM) in combination with scanning transmission electron microscope (STEM)-energy dispersive x-ray spectroscopy (EDS) mapping. Significant differences were seen in microstructural evolution between the two different microstructures that were irradiated under similar conditions. Nucleation and growth of < c >-component loops and their dependence on the alloying elements are a major focus of the current investigation. It was observed that the < c >-component loops nucleate readily at 100, 300, and 400 degrees C after a threshold incubation dose (TID), which varies with irradiation temperature and the state of alloying elements. It was found that the TID for the formation of < c >-component loops increases with decrease in irradiation temperature. Alloying elements that are present in the form of SPPs increase the TID compared to when they are in the beta phase solid solution. Dose and temperature dependence of loop size and density are presented. Radiation induced redistribution and clustering of alloying elements (Sn, Mo, and Fe) have been observed and related to the formation of < c >-component loops. It has been shown that at the higher temperature tests, irradiation induced dissolution of precipitates occurs whereas irradiation induced

Low temperature neutron irradiation effects on microstructure and tensile properties of molybdenum

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Meimei; Eldrup, M.; Byun, Thak Sang

2008-01-01

Polycrystalline molybdenum was irradiated in the hydraulic tube facility at the High Flux Isotope Reactor to doses ranging from 7.2 x 10{sup -5} to 0.28 dpa at {approx} 80 C. As-irradiated microstructure was characterized by room-temperature electrical resistivity measurements, transmission electron microscopy (TEM) and positron annihilation spectroscopy (PAS). Tensile tests were carried out between -50 and 100 C over the strain rate range 1 x 10{sup -5} to 1 x 10{sup -2} s{sup -1}. Fractography was performed by scanning electron microscopy (SEM), and the deformation microstructure was examined by TEM after tensile testing. Irradiation-induced defects became visible by TEM atmore » {approx}0.001 dpa. Both their density and mean size increased with increasing dose. Submicroscopic three-dimensional cavities were detected by PAS even at {approx}0.0001 dpa. The cavity density increased with increasing dose, while their mean size and size distribution was relatively insensitive to neutron dose. It is suggested that the formation of visible dislocation loops was predominantly a nucleation and growth process, while in-cascade vacancy clustering may be significant in Mo. Neutron irradiation reduced the temperature and strain rate dependence of the yield stress, leading to radiation softening in Mo at lower doses. Irradiation had practically no influence on the magnitude and the temperature and strain rate dependence of the plastic instability stress.« less

Database on Performance of Neutron Irradiated FeCrAl Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Briggs, Samuel A.; Littrell, Ken

The present report summarizes and discusses the database on radiation tolerance for Generation I, Generation II, and commercial FeCrAl alloys. This database has been built upon mechanical testing and microstructural characterization on selected alloys irradiated within the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) up to doses of 13.8 dpa at temperatures ranging from 200°C to 550°C. The structure and performance of these irradiated alloys were characterized using advanced microstructural characterization techniques and mechanical testing. The primary objective of developing this database is to enhance the rapid development of a mechanistic understanding on the radiation tolerancemore » of FeCrAl alloys, thereby enabling informed decisions on the optimization of composition and microstructure of FeCrAl alloys for application as an accident tolerant fuel (ATF) cladding. This report is structured to provide a brief summary of critical results related to the database on radiation tolerance of FeCrAl alloys.« less

PROCESS FOR DISSOLVING BINARY URANIUM-ZIRCONIUM OR ZIRCONIUM-BASE ALLOYS

DOEpatents

Jonke, A.A.; Barghusen, J.J.; Levitz, N.M.

1962-08-14

A process of dissolving uranium-- zirconium and zircaloy alloys, e.g. jackets of fuel elements, with an anhydrous hydrogen fluoride containing from 10 to 32% by weight of hydrogen chloride at between 400 and 450 deg C., preferably while in contact with a fluidized inert powder, such as calcium fluoride is described. (AEC)

Atomistic modeling of high temperature uranium-zirconium alloy structure and thermodynamics

NASA Astrophysics Data System (ADS)

Moore, A. P.; Beeler, B.; Deo, C.; Baskes, M. I.; Okuniewski, M. A.

2015-12-01

A semi-empirical Modified Embedded Atom Method (MEAM) potential is developed for application to the high temperature body-centered-cubic uranium-zirconium alloy (γ-U-Zr) phase and employed with molecular dynamics (MD) simulations to investigate the high temperature thermo-physical properties of U-Zr alloys. Uranium-rich U-Zr alloys (e.g. U-10Zr) have been tested and qualified for use as metallic nuclear fuel in U.S. fast reactors such as the Integral Fast Reactor and the Experimental Breeder Reactors, and are a common sub-system of ternary metallic alloys like U-Pu-Zr and U-Zr-Nb. The potential was constructed to ensure that basic properties (e.g., elastic constants, bulk modulus, and formation energies) were in agreement with first principles calculations and experimental results. After which, slight adjustments were made to the potential to fit the known thermal properties and thermodynamics of the system. The potentials successfully reproduce the experimental melting point, enthalpy of fusion, volume change upon melting, thermal expansion, and the heat capacity of pure U and Zr. Simulations of the U-Zr system are found to be in good agreement with experimental thermal expansion values, Vegard's law for the lattice constants, and the experimental enthalpy of mixing. This is the first simulation to reproduce the experimental thermodynamics of the high temperature γ-U-Zr metallic alloy system. The MEAM potential is then used to explore thermodynamics properties of the high temperature U-Zr system including the constant volume heat capacity, isothermal compressibility, adiabatic index, and the Grüneisen parameters.

Ion irradiation induced defect evolution in Ni and Ni-based FCC equiatomic binary alloys

DOE PAGES

Jin, Ke; Zhang, Yanwen; Bei, Hongbin

2015-09-09

In order to explore the chemical effects on radiation response of alloys with multi-principal elements, defect evolution under Au ion irradiation was investigated in the elemental Ni, equiatomic NiCo and NiFe alloys. Single crystals were successfully grown in an optical floating zone furnace and their (100) surfaces were irradiated with 3 MeV Au ions at fluences ranging from 1 × 10 13 to 5 × 10 15 ions cm –2 at room temperature. The irradiation-induced defect evolution was analyzed by using ion channeling technique. Experiment shows that NiFe is more irradiation-resistant than NiCo and pure Ni at low fluences. Withmore » continuously increasing the ion fluences, damage level is eventually saturated for all materials but at different dose levels. The saturation level in pure Ni appears at relatively lower irradiation fluence than the alloys, suggesting that damage accumulation slows down in the alloys. Here, under high-fluence irradiations, pure Ni has wider damage ranges than the alloys, indicating that defects in pure Ni have high mobility.« less

Ion irradiation testing and characterization of FeCrAl candidate alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anderoglu, Osman; Aydogan, Eda; Maloy, Stuart Andrew

2014-10-29

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commerciallymore » available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.« less

An interdiffusional model for prediction of the interaction layer growth in the system uranium molybdenum/aluminum

NASA Astrophysics Data System (ADS)

Soba, A.; Denis, A.

2007-03-01

The codes PLACA and DPLACA, elaborated in this working group, simulate the behavior of a plate-type fuel containing in its core a foil of monolithic or dispersed fissile material, respectively, under normal operation conditions of a research reactor. Dispersion fuels usually consist of ceramic particles of a uranium compound in a high thermal conductivity matrix. The use of particles of a U-Mo alloy in a matrix of Al requires especially devoted subroutines able to simulate the growth of the interaction layer that develops between the particles and the matrix. A model is presented in this work that gives account of these particular phenomena. It is based on the assumption that diffusion of U and Al through the layer is the rate-determining step. Two moving interfaces separate the growing reaction layer from the original phases. The kinetics of these boundaries are solved as Stefan problems. In order to test the model and the associated code, some previous, simpler problems corresponding to similar systems for which analytical solutions or experimental data are known were simulated. Experiments performed with planar U-Mo/Al diffusion couples are reported in the literature, which purpose is to obtain information on the system parameters. These experiments were simulated with PLACA. Results of experiments performed with U-Mo particles disperse in Al either without or with irradiation, published in the open literature were simulated with DPLACA. A satisfactory prediction of the whole reaction layer thickness and of the individual fractions corresponding to alloy and matrix consumption was obtained.

High Temperature Irradiation-Resistant Thermocouple Performance Improvements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Joshua Daw; Joy Rempe; Darrell Knudson

2009-04-01

Traditional methods for measuring temperature in-pile degrade at temperatures above 1100 ºC. To address this instrumentation need, the Idaho National Laboratory (INL) developed and evaluated the performance of a high temperature irradiation-resistant thermocouple (HTIR-TC) that contains alloys of molybdenum and niobium. Data from high temperature (up to 1500 ºC) long duration (up to 4000 hours) tests and on-going irradiations at INL’s Advanced Test Reactor demonstrate the superiority of these sensors to commercially-available thermocouples. However, several options have been identified that could further enhance their reliability, reduce their production costs, and allow their use in a wider range of operating conditions.more » This paper presents results from on-going Idaho National Laboratory (INL)/University of Idaho (UI) efforts to investigate options to improve HTIR-TC ductility, reliability, and resolution by investigating specially-formulated alloys of molybdenum and niobium and alternate diameter thermoelements (wires). In addition, on-going efforts to evaluate alternate fabrication approaches, such as drawn and loose assembly techniques will be discussed. Efforts to reduce HTIR-TC fabrication costs, such as the use of less expensive extension cable will also be presented. Finally, customized HTIR-TC designs developed for specific customer needs will be summarized to emphasize the varied conditions under which these sensors may be used.« less

Uptake of trace elements and radionuclides from uranium mill tailings by four-wing saltbush (Atriplex canescens) and alkali sacaton (Sporobolus airoides). [Radium 226; Uranium; Molybdenum; Selenium; Vanadium; Astatine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dreesen, D.R.; Marple, M.L.

1979-01-01

A greenhouse experiment was performed to determine the uptake of trace elements and radionuclides from uranium mill tailings by native plant species. Four-wing saltbush and alkali sacaton were grown in alkaline tailings covered with soil and in soil alone as controls. The tailings material was highly enriched in Ra-226, Mo, U, Se, V, and As compared with three local soils. The shrub grown in tailings had elevated concentrations of Mo, Se, Ra-226, U, As, and Na compared with the controls. Alkali sacaton contained high concentrations of Mo, Se, Ra-226, and Ni when grown on tailings. Molybdenum and selenium concentrations inmore » plants grown in tailings are above levels reported to be toxic to grazing animals. These results indicate that the bioavailability of Mo and Se in alkaline environments makes these elements among the most hazardous contaminants present in uranium mill wastes.« less

Characterization of high energy Xe ion irradiation effects in single crystal molybdenum with depth-resolved synchrotron microbeam diffraction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yun, Di; Miao, Yinbin; Xu, Ruqing

2016-04-01

Microbeam X-ray diffraction experiments were conducted at beam line 34-ID of the Advanced Photon Source (APS) on fission fragment energy Xe heavy ion irradiated single crystal Molybdenum (Mo). Lattice strain measurements were obtained with a depth resolution of 0.7 mu m, which is critical in resolving the peculiar heterogeneity of irradiation damage associated with heavy ion irradiation. Q-space diffraction peak shift measurements were correlated with lattice strain induced by the ion irradiations. Transmission electron microscopy (TEM) characterizations were performed on the as-irradiated materials as well. Nanometer sized Xe bubble microstructures were observed via TEM. Molecular Dynamics (MD) simulations were performedmore » to help interpret the lattice strain measurement results from the experiment. This study showed that the irradiation effects by fission fragment energy Xe ion irradiations can be collaboratively understood with the depth resolved X-ray diffraction and TEM measurements under the assistance of MD simulations. (c) 2015 Elsevier B.V. All rights reserved.« less

Distribution of selenium, molybdenum and uranium in sediment cores from the Colorado River delta, Baja California, Mexico.

PubMed

Orozco-Durán, A; Daesslé, L W; Gutiérrez-Galindo, E A; Muñoz-Barbosa, A

2012-01-01

The distribution of selenium, molybdenum and uranium was studied in ~1.5 m sediment cores from the Colorado River delta, at the Colorado (CR) and Hardy (HR) riverbeds. Core HR2 showed highest Se, Mo and U concentrations at its bottom (2.3, 0.95 and 1.8 μg g(-1)) within a sandy-silt layer deposited prior to dam construction. In CR5 the highest concentrations of these elements (0.9, 1.4 and 1.7 μg g(-1) respectively) were located at the top of the core within a surface layer enriched in organic carbon. A few samples from HR2 had Se above the probable toxic effect level guidelines.

COATING URANIUM FROM CARBONYLS

DOEpatents

Gurinsky, D.H.; Storrs, S.S.

1959-07-14

Methods are described for making adherent corrosion resistant coatings on uranium metal. According to the invention, the uranium metal is heated in the presence of an organometallic compound such as the carbonyls of nickel, molybdenum, chromium, niobium, and tungsten at a temperature sufficient to decompose the metal carbonyl and dry plate the resultant free metal on the surface of the uranium metal body. The metal coated body is then further heated at a higher temperature to thermally diffuse the coating metal within the uranium bcdy.

Influence of a doping by Al stainless steel on kinetics and character of interaction with the metallic nuclear fuel

NASA Astrophysics Data System (ADS)

Nikitin, S. N.; Shornikov, D. P.; Tarasov, B. A.; Baranov, V. G.

2016-04-01

Metallic nuclear fuel is a perspective kind of fuel for fast reactors. In this paper we conducted a study of the interaction between uranium-molybdenum alloy and ferritic- martensitic steels with additions of aluminum at a temperature of 700 ° C for 25 hours. The rate constants of the interaction layer growth at 700 °C is about 2.8.10-14 m2/s. It is established that doping Al stainless steel leads to decrease in interaction with uranium-molybdenum alloys. The phase composition of the interaction layer is determined.

Silver and palladium alloy nanoparticle catalysts: reductive coupling of nitrobenzene through light irradiation.

PubMed

Peiris, Sunari; Sarina, Sarina; Han, Chenhui; Xiao, Qi; Zhu, Huai-Yong

2017-08-15

Silver-palladium (Ag-Pd) alloy nanoparticles strongly absorb visible light and exhibit significantly higher photocatalytic activity compared to both pure palladium (Pd) and silver (Ag) nanoparticles. Photocatalysts of Ag-Pd alloy nanoparticles on ZrO 2 and Al 2 O 3 supports are developed to catalyze the nitroaromatic coupling to the corresponding azo compounds under visible light irradiation. Ag-Pd alloy NP/ZrO 2 exhibited the highest photocatalytic activity for nitrobenzene coupling to azobenzene (yield of ∼80% in 3 hours). The photocatalytic efficiency could be optimized by altering the Ag : Pd ratio of the alloy nanoparticles, irradiation light intensity, temperature and wavelength. The rate of the reaction depends on the population and energy of the excited electrons, which can be improved by increasing the light intensity or by using a shorter wavelength. The knowledge developed in this study may inspire further studies on Ag alloy photocatalysts and organic syntheses using Ag-Pd nanoparticle catalysts driven under visible light Irradiation.

Fabrication, strength and oxidation of molybdenum-silicon-boron alloys from reaction synthesis

NASA Astrophysics Data System (ADS)

Middlemas, Michael Robert

Mo-Si-B alloys are a leading candidate for the next generation of jet turbine engine blades and have the potential to raise the operating temperatures by 300-400°C, which would dramatically increase power and efficiency. The alloys of interest are a three-phase mixture of the molybdenum solid solution (Moss) and two intermetallic phases, Mo3Si (A15) and Mo5SiB2 (T2). A novel powder metallurgical method was developed which uses the reaction of molybdenum, silicon nitride (Si3N4) and boron nitride (BN) powders to synthesize a fine dispersion of the intermetallic phases in a Moss matrix. The covalent nitrides are stable in oxidizing environments up to 1000ºC, allowing for fine particle processing without the formation of silicon and boron oxides. The process developed uses standard powder processing techniques to create Mo-Si-B alloys in a less complex and expensive manner than previously demonstrated. The formation of the intermetallic phases was examined by thermo-gravimetric analysis and x-ray diffraction. The start of the reactions to form the T2 and A15 phases were observed at 1140°C and 1193°C and the reactions have been demonstrated to be complete in as little as two hours at 1300°C. This powder metallurgy approach yields a fine dispersion of intermetallics in the Moss matrix, with average grain sizes of 2-4mum. Densities up to 95% of theoretical were attained from pressureless sintering at 1600°C and full theoretical density was achieved by hot-isostatic pressing (HIP). Low temperature sintering and HIPing was attempted to limit grain growth and to reduce the equilibrium silicon concentration in the Moss matrix. Sintering and HIPing at 1300°C reduced the grain sizes of all three phases by over a factor of two. Powder metallurgy provides an opportunity for microstructure control through changes in raw materials and processing parameters. Microstructure examination by electron back-scatter diffraction (EBSD) imaging was used to precisely define the

Stress corrosion cracking behavior of irradiated model austenitic stainless steel alloys.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, H. M.; Karlsen, T. M.; Ruther, W. E.

Slow-strain-rate tensile tests (SSRTs) and posttest fractographic analyses by scanning electron microscopy were conducted on 16 austenitic stainless steel (SS) alloys that were irradiated at 289 C in He. After irradiation to {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2} and {approx}0.9 x 10{sup 21} n{center_dot}cm{sup {minus}2} (E >1 MeV), significant heat-to-heat variations in the degree of intergranular and transgranular stress corrosion cracking (IGSCC and TGSCC) were observed. Following irradiation to a fluence of {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2}, a high-purity laboratory heat of Type 316L SS (Si {approx} 0.024 wt%) exhibited the highest susceptibility to IGSCC. The other 15 alloysmore » exhibited negligible susceptibility to IGSCC at this low fluence. The percentage of TGSCC on the fracture surfaces of SSRT specimens of the 16 alloys at {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2} (E > 1 MeV) could be correlated well with N and Si concentrations; all alloys that contained <0.01 wt.% N and <1.0 wt. % Si were susceptible, whereas all alloys that contained >0.01 wt.% N or >1.0 wt.% Si were relatively resistant. High concentrations of Cr were beneficial. Alloys that contain <15.5 wt.% Cr exhibited greater percentages of TGSCC and IGSCC than those alloys with {approx}18 wt.% Cr, whereas an alloy that contains >21 wt.% Cr exhibited less susceptibility than the lower-Cr alloys under similar conditions.« less

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation - Non-destructive analysis of the AFIP-1 fuel plates

NASA Astrophysics Data System (ADS)

Wachs, D. M.; Robinson, A. B.; Rice, F. J.; Kraft, N. C.; Taylor, S. C.; Lillo, M.; Woolstenhulme, N.; Roth, G. A.

2016-08-01

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008-2009. The irradiation conditions were: ∼250 W/cm2 peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm3 peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

Radiation damage studies of ion-irradiated low-activation developmental martensitic steel alloys for fusion applications

NASA Astrophysics Data System (ADS)

Mazey, D. J.; Hanks, W.; Lurcook, O. K.

1990-09-01

Five martensitic, nominally 9 and 11% Cr-W-V-Mn-Ta stainless steels which have been developed as low-activation alloys for fusion-reactor structural applications have been irradiated with 52 MeV Cr 6+ ions to 20 dpa at 475°C in the Harwell Variable Energy Cyclotron (VEC). Four of the alloys contained additions of 0.1 wt% Ta and these had been shown in prior tests to have mechanical properties comparable with the conventional FV 448 alloy. Examinations by TEM showed that irradiation-induced precipitates were present on a fine-scale in all of the alloys. These comprised Cr-rich lath-like defects in the 9Cr, Ta-free alloy; small Cr-rich particles in the 9Cr-3W-0.1Ta alloy and Cr-rich planar precipitates in the remaining alloys. Little or no irradiation-induced cavitation was observed. The other important irradiation-induced response was in the dislocation structure in the Ta-containing alloys which comprised an extensive rafted array of elongated a <100> type dislocation loops having major axes aligned in <100> directions. A significant fraction of the presumed a <100> loops contained stacking-fault fringes and analysis suggested that these were Cr 2N or Fe 4N nitride phase which it is known can form on {001} habit planes. Such nitrides are observed frequently under thermal-annealing conditions in ferritic steels, but less frequently under irradiation. Their formation in relation to the void swelling resistance of ferritic-martensitic alloys is discussed.

Irradiation-enhanced α' precipitation in model FeCrAl alloys

DOE PAGES

Edmondson, Philip D.; Briggs, Samuel A.; Yamamoto, Yukinori; ...

2016-02-17

We have irradiated the model FeCrAl alloys with varying compositions (Fe(10–18)Cr(10–6)Al at.%) with a neutron at ~ 320 to damage levels of ~ 7 displacements per atom (dpa) to investigate the compositional influence on the formation of irradiation-induced Cr-rich α' precipitates using atom probe tomography. In all alloys, significant number densities of these precipitates were observed. Cluster compositions were investigated and it was found that the average cluster Cr content ranged between 51.1 and 62.5 at.% dependent on initial compositions. Furthermore, this is significantly lower than the Cr-content of α' in binary FeCr alloys. As a result, significant partitioning ofmore » the Al from the α' precipitates was also observed.« less

Early breakthrough of molybdenum and uranium in a permeable reactive barrier.

PubMed

Morrison, Stan J; Mushovic, Paul S; Niesen, Preston L

2006-03-15

A permeable reactive barrier (PRB) using zerovalent iron (ZVI) was installed at a site near Cañon City, CO, to treat molybdenum (Mo) and uranium (U) in groundwater. The PRB initially decreased Mo concentrations from about 4.8 to less than 0.1 mg/L; however, Mo concentrations in the ZVI increased to 2.0 mg/L after about 250 days and continued to increase until concentrations in the ZVI were about 4 times higherthan in the influent groundwater. Concentrations of U were reduced from 1.0 to less than 0.02 mg/L during the same period. Investigations of solid-phase samples indicate that (1) calcium carbonate, iron oxide, and sulfide minerals had precipitated in pores of the ZVI; (2) U and Mo were concentrated in the upgradient 5.1 cm of the ZVI; and (3) calcium was present throughout the ZVI accounting for up to 20.5% of the initial porosity. Results of a column test indicated that the ZVI from the PRB was still reactive for removing Mo and that removal rates were dependenton residence time and pH. The chemical evolution of the PRB is explained in four stages that present a progression from porous media flow through preferential flow and, finally, complete bypass of the ZVI.

A phase field model for segregation and precipitation induced by irradiation in alloys

NASA Astrophysics Data System (ADS)

Badillo, A.; Bellon, P.; Averback, R. S.

2015-04-01

A phase field model is introduced to model the evolution of multicomponent alloys under irradiation, including radiation-induced segregation and precipitation. The thermodynamic and kinetic components of this model are derived using a mean-field model. The mobility coefficient and the contribution of chemical heterogeneity to free energy are rescaled by the cell size used in the phase field model, yielding microstructural evolutions that are independent of the cell size. A new treatment is proposed for point defect clusters, using a mixed discrete-continuous approach to capture the stochastic character of defect cluster production in displacement cascades, while retaining the efficient modeling of the fate of these clusters using diffusion equations. The model is tested on unary and binary alloy systems using two-dimensional simulations. In a unary system, the evolution of point defects under irradiation is studied in the presence of defect clusters, either pre-existing ones or those created by irradiation, and compared with rate theory calculations. Binary alloys with zero and positive heats of mixing are then studied to investigate the effect of point defect clustering on radiation-induced segregation and precipitation in undersaturated solid solutions. Lastly, irradiation conditions and alloy parameters leading to irradiation-induced homogeneous precipitation are investigated. The results are discussed in the context of experimental results reported for Ni-Si and Al-Zn undersaturated solid solutions subjected to irradiation.

Potential annealing treatments for tailoring the starting microstructure of low-enriched U-Mo dispersion fuels to optimize performance during irradiation

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Woolstenhulme, Nicolas E.; Ewh, Ashley

2011-12-01

Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525 °C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475 °C.

Ternary cobalt-molybdenum-zirconium coatings for alternative energies

NASA Astrophysics Data System (ADS)

Yar-Mukhamedova, Gulmira; Ved', Maryna; Sakhnenko, Nikolay; Koziar, Maryna

2017-11-01

Consistent patterns for electrodeposition of Co-Mo-Zr coatings from polyligand citrate-pyrophosphate bath were investigated. The effect of both current density amplitude and pulse on/off time on the quality, composition and surface morphology of the galvanic alloys were determined. It was established the coating Co-Mo-Zr enrichment by molybdenum with current density increasing up to 8 A dm-2 as well as the rising of pulse time and pause duration promotes the content of molybdenum because of subsequent chemical reduction of its intermediate oxides by hydrogen ad-atoms. It was found that the content of the alloying metals in the coating Co-Mo-Zr depends on the current density and on/off times extremely and maximum Mo and Zr content corresponds to the current density interval 4-6 A dm-2, on-/off-time 2-10 ms. Chemical resistance of binary and ternary coatings based on cobalt is caused by the increased tendency to passivity and high resistance to pitting corrosion in the presence of molybdenum and zirconium, as well as the acid nature of their oxides. Binary coating with molybdenum content not less than 20 at.% and ternary ones with zirconium content in terms of corrosion deep index are in a group ;very proof;. It was shown that Co-Mo-Zr alloys exhibits the greatest level of catalytic properties as cathode material for hydrogen electrolytic production from acidic media which is not inferior a platinum electrode. The deposits Co-Mo-Zr with zirconium content 2-4 at.% demonstrate high catalytic properties in the carbon(II) oxide conversion. This confirms the efficiency of materials as catalysts for the gaseous wastes purification and gives the reason to recommend them as catalysts for red-ox processes activating by oxygen as well as electrode materials for red-ox batteries.

High temperature alloy

NASA Technical Reports Server (NTRS)

Frank, R. G.; Semmel, J. W., Jr.

1968-01-01

Molybdenum is substituted for tungsten on an atomic basis in a cobalt-based alloy, S-1, thus enabling the alloy to be formed into various mill products, such as tubing and steels. The alloy is weldable, has good high temperature strength and is not subject to embrittlement produced by high temperature aging.

Effect of solute atom concentration on vacancy cluster formation in neutron-irradiated Ni alloys

NASA Astrophysics Data System (ADS)

Sato, Koichi; Itoh, Daiki; Yoshiie, Toshimasa; Xu, Qiu; Taniguchi, Akihiro; Toyama, Takeshi

2011-10-01

The dependence of microstructural evolution on solute atom concentration in Ni alloys was investigated by positron annihilation lifetime measurements. The positron annihilation lifetimes in pure Ni, Ni-0.05 at.%Si, Ni-0.05 at.%Sn, Ni-Cu, and Ni-Ge alloys were about 400 ps even at a low irradiation dose of 3 × 10 -4 dpa, indicating the presence of microvoids in these alloys. The size of vacancy clusters in Ni-Si and Ni-Sn alloys decreased with an increase in the solute atom concentration at irradiation doses less than 0.1 dpa; vacancy clusters started to grow at an irradiation dose of about 0.1 dpa. In Ni-2 at.%Si, irradiation-induced segregation was detected by positron annihilation coincidence Doppler broadening measurements. This segregation suppressed one-dimensional (1-D) motion of the interstitial clusters and promoted mutual annihilation of point defects. The frequency and mean free path of the 1-D motion depended on the solute atom concentration and the amount of segregation.

JACKETING URANIUM

DOEpatents

Saller, H.A.; Keeler, J.R.

1959-07-14

The bonding to uranium of sheathing of iron or cobalt, or nickel, or alloys thereof is described. The bonding is accomplished by electro-depositing both surfaces to be joined with a coating of silver and amalgamating or alloying the silver layer with mercury or indium. Then the silver alloy is homogenized by exerting pressure on an assembly of the uranium core and the metal jacket, reducing the area of assembly and heating the assembly to homogenize by diffusion.

Solid solution strengthened duct and cladding alloy D9-B1

DOEpatents

Korenko, Michael K.

1983-01-01

A modified AISI type 316 stainless steel is described for use in an atmosphere where the alloy will be subject to neutron irradiation. The alloy is characterized by its phase stability in both the annealed as well as cold work condition and above all by its superior resistance to radiation induced swelling. Graphical data is included to demonstrate the superior swelling resistance of the alloy which contains from about 0.5% to 2.2% manganese, from about 0.7% to about 1.1% silicon, from about 12.5% to 14% chromium, from about 14.5% to about 16.5% nickel, from about 1.2% to about 1.6% molybdenum, from 0.15% to 0.30% titanium, from 0.02% to 0.08% zirconium, and the balance iron with incidental impurities.

URANIUM COMPOSITIONS

DOEpatents

Allen, N.P.; Grogan, J.D.

1959-05-12

This patent relates to high purity uranium alloys characterized by improved stability to thermal cycling and low thermal neutron absorption. The high purity uranium alloy contains less than 0.1 per cent by weight in total amount of any ore or more of the elements such as aluminum, silicon, phosphorous, tin, lead, bismuth, niobium, and zinc.

Effects of ultraviolet irradiation on bonding strength between Co-Cr alloy and citric acid-crosslinked gelatin matrix.

PubMed

Inoue, Motoki; Sasaki, Makoto; Katada, Yasuyuki; Taguchi, Tetsushi

2014-02-01

Novel techniques for creating a strong bond between polymeric matrices and biometals are required. We immobilized polymeric matrices on the surface of biometal for drug-eluting stents through covalent bond. We performed to improve the bonding strength between a cobalt-chromium alloy and a citric acid-crosslinked gelatin matrix by ultraviolet irradiation on the surface of cobalt-chromium alloy. The ultraviolet irradiation effectively generated hydroxyl groups on the surface of the alloy. The bonding strength between the gelatin matrix and the alloy before ultraviolet irradiation was 0.38 ± 0.02 MPa, whereas it increased to 0.48 ± 0.02 MPa after ultraviolet irradiation. Surface analysis showed that the citric acid derivatives occurred on the surface of the cobalt-chromium alloy through ester bond. Therefore, ester bond formation between the citric acid derivatives active esters and the hydroxyl groups on the cobalt-chromium alloy contributed to the enhanced bonding strength. Ultraviolet irradiation and subsequent immobilization of a gelatin matrix using citric acid derivatives is thus an effective way to functionalize biometal surfaces.

Effect of neutron irradiation on the thermoelectric properties of SiGe alloys

NASA Technical Reports Server (NTRS)

Vandersande, Jan W.; Mccormack, Joe; Zoltan, Andy; Farmer, John

1990-01-01

Zone-leveled and hot-pressed n- and p-type Si80Ge20 alloys were irradiated with neutrons to a fluence of 4 x 1018 n/sq cm and to a fluence of 5.4 x 1019 n/sq cm at a temperature of approximately 200-300 C. The effect of neutron irradiation on the thermoelectric properties of these alloys was evaluated. The carrier concentration and mobility (and hence the resistivity) were measured at room temperature while the thermal diffusivity was measured at 177-192 C both before and after the irradiation and after each subsequent 2-h heat treatment at 350 C, 600, and 1000 C. The irradiation increased the resistivity significantly, but the thermal conductivity decreased only by about 10-15 percent. This tends to indicate that the radiation produced only small defects (single pairs and small vacancy chains). The samples all returned to almost exactly their preirradiation state after the 1000 C anneal. This indicates that SiGe alloys can be operated in this neutron fluence at high temperatures without a degradation of thermoelectric properties.

Method for fabricating laminated uranium composites

DOEpatents

Chapman, L.R.

1983-08-03

The present invention is directed to a process for fabricating laminated composites of uranium or uranium alloys and at least one other metal or alloy. The laminated composites are fabricated by forming a casting of the molten uranium with the other metal or alloy which is selectively positioned in the casting and then hot-rolling the casting into a laminated plate in or around which the casting components are metallurgically bonded to one another to form the composite. The process of the present invention provides strong metallurgical bonds between the laminate components primarily since the bond disrupting surface oxides on the uranium or uranium alloy float to the surface of the casting to effectively remove the oxides from the bonding surfaces of the components.

PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

DOEpatents

Moore, R.H.

1962-10-01

A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

Plasmachemical synthesis and evaluation of the thermal conductivity of metal-oxide compounds "Molybdenum-uranium dioxide"

NASA Astrophysics Data System (ADS)

Kotelnikova, Alexandra A.; Karengin, Alexander G.; Mendoza, Orlando

2018-03-01

The article represents possibility to apply oxidative and reducing plasma for plasma-chemical synthesis of metal-oxide compounds «Mo‒UO2» from water-salt mixtures «molybdic acid‒uranyl nitrate» and «molybdic acid‒ uranyl acetate». The composition of water-salt mixture was calculated and the conditions ensuring plasma-chemical synthesis of «Mo‒UO2» compounds were determined. Calculations were carried out at atmospheric pressure over a wide range of temperatures (300-4000 K), with the use of various plasma coolants (air, hydrogen). The heat conductivity coefficients of metal-oxide compounds «Mo‒UO2» consisting of continuous component (molybdenum matrix) are calculated. Inclusions from ceramics in the form of uranium dioxide were ordered in the matrix. Particular attention is paid to methods for calculating the coefficients of thermal conductivity of these compounds with the use of different models. Calculated results were compared with the experimental data.

Role of Si on the Diffusional Interactions Between U-Mo and Al-Si Alloys at 823 K (550 °C)

NASA Astrophysics Data System (ADS)

Perez, Emmanuel; Sohn, Yong-Ho; Keiser, Dennis D.

2013-01-01

U-Mo dispersions in Al-alloy matrix and monolithic fuels encased in Al-alloy are under development to fulfill the requirements for research and test reactors to use low-enriched molybdenum stabilized uranium alloy fuels. Significant interaction takes place between the U-Mo fuel and Al during manufacturing and in-reactor irradiation. The interaction products are Al-rich phases with physical and thermal characteristics that adversely affect fuel performance and result in premature failure. Detailed analysis of the interdiffusion and microstructural development of this system was carried through diffusion couples consisting of U-7 wt pct Mo, U-10 wt pct Mo and U-12 wt pct Mo in contact with pure Al, Al-2 wt pct Si, and Al-5 wt pct Si, annealed at 823 K (550 °C) for 1, 5 and 20 hours. Scanning electron microscopy and transmission electron microscopy were employed for the analysis. Diffusion couples consisting of U-Mo in contact with pure Al contained UAl3, UAl4, U6Mo4Al43, and UMo2Al20 phases. Additions of Si to the Al significantly reduced the thickness of the interdiffusion zone. The interdiffusion zones developed Al- and Si-enriched regions, whose locations and size depended on the Si and Mo concentrations in the terminal alloys. In these couples, the (U,Mo)(Al,Si)3 phase was observed throughout the interdiffusion zone, and the U6Mo4Al43 and UMo2Al20 phases were observed only where the Si concentrations were low.

Uranium-molybdenum nuclear fuel plates behaviour under heavy ion irradiation: An X-ray diffraction analysis

NASA Astrophysics Data System (ADS)

Palancher, H.; Wieschalla, N.; Martin, P.; Tucoulou, R.; Sabathier, C.; Petry, W.; Berar, J.-F.; Valot, C.; Dubois, S.

2009-03-01

Heavy ion irradiation has been proposed for discriminating UMo/Al specimens which are good candidates for research reactor fuels. Two UMo/Al dispersed fuels (U-7 wt%Mo/Al and U-10 wt%Mo/Al) have been irradiated with a 80 MeV 127I beam up to an ion fluence of 2 × 1017 cm-2. Microscopy and mainly X-ray diffraction using large and micrometer sized beams have enabled to characterize the grown interaction layer: UAl3 appears to be the only produced crystallized phase. The presence of an amorphous additional phase can however not be excluded. These results are in good agreement with characterizations performed on in-pile irradiated fuels and encourage new studies with heavy ion irradiation.

Amorphous metal alloy

DOEpatents

Wang, R.; Merz, M.D.

1980-04-09

Amorphous metal alloys of the iron-chromium and nickel-chromium type have excellent corrosion resistance and high temperature stability and are suitable for use as a protective coating on less corrosion resistant substrates. The alloys are stabilized in the amorphous state by one or more elements of titanium, zirconium, hafnium, niobium, tantalum, molybdenum, and tungsten. The alloy is preferably prepared by sputter deposition.

Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

DOEpatents

McLean, II, William; Miller, Philip E.

1997-01-01

A method for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction.

Damage accumulation in ion-irradiated Ni-based concentrated solid-solution alloys

DOE PAGES

Ullah, Mohammad W.; Aidhy, Dilpuneet S.; Zhang, Yanwen; ...

2016-03-05

We investigate Irradiation-induced damage accumulation in Ni 0.8Fe 0.2 and Ni 0.8Cr 0.2 alloys by using molecular dynamics simulations to assess possible enhanced radiation-resistance in these face-centered cubic (fcc), single-phase, concentrated solid-solution alloys, as compared with pure fcc Ni.

The irradiation-induced microstructural development and the role of γ' on void formation in Ni-based alloys

NASA Astrophysics Data System (ADS)

Kato, Takahiko; Nakata, Kiyotomo; Masaoka, Isao; Takahashi, Heishichiro; Takeyama, Taro; Ohnuki, Soumei; Osanai, Hisashi

1984-05-01

The microstructural development for Inconel X-750, N1-13 at%A1, and Ni-11.5 at%Si alloys during irradiation was investigated. These alloys were previously heat-treated at temperatures of 723-1073 K, and γ' precipitates were produced. Irradiation was performed in a high voltage electron microscope (1000 kV) in the temperature range 673-823 K. In the case of solution-treated Inconel, interstitial dislocation loops were formed initially, while voids were nucleated after longer times. When the Inconel specimen containing a high number density of small γ' was irradiated, dislocation loops were formed in both the matrix and precipitate-matrix interface. The loops formed on the interface scarcely grew during irradiation. On the other hand, for the Ni-Al alloy fine γ' nucleated during irradiation, the large γ' precipitated by pre-aging, dissolved. A similar resolution process was also observed in Ni-Si alloy. Furthermore, in the Ni-Si alloy precipitates of γ' formed preferentially at interstitial dislocation loops and both specimen surfaces.

Studies on the reactive melt infiltration of silicon and silicon-molybdenum alloys in porous carbon

NASA Technical Reports Server (NTRS)

Singh, M.; Behrendt, D. R.

1992-01-01

Investigations on the reactive melt infiltration of silicon and silicon-1.7 and 3.2 at percent molybdenum alloys into porous carbon preforms have been carried out by process modeling, differential thermal analysis (DTA) and melt infiltration experiments. These results indicate that the initial pore volume fraction of the porous carbon preform is a critical parameter in determining the final composition of the raction-formed silicon carbide and other residual phases. The pore size of the carbon preform is very detrimental to the exotherm temperatures due to liquid silicon-carbon reactions encountered during the reactive melt infiltration process. A possible mechanism for the liquid silicon-porous (glassy) carbon reaction has been proposed. The composition and microstructure of the reaction-formed silicon carbide has been discussed in terms of carbon preform microstructures, infiltration materials, and temperatures.

Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

DOEpatents

McLean, W. II; Miller, P.E.

1997-12-16

A method is described for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction. 3 figs.

Examination of T-111 clad uranium nitride fuel pins irradiated up to 13,000 hours at a clad temperature of 990 C

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.

1973-01-01

The examination of 27 fuel pins irradiated for up to 13,000 hours at 990 C is described. The fuel pin clad was a tantalum alloy with uranium nitride as the nuclear fuel. Two nominal fuel pin diameters were tested with a maximum burnup of 2.34 atom percent. Twenty-two fuel pins were tested for fission gas leaks; thirteen pins leaked. Clad ductility tests indicated clad embrittlement. The embrittlement is attributed to hydrogen from an n,p reaction in the fuel. Fuel swelling was burnup dependent, and the amount of fission gas release was low, generally less than 0.5 percent. No incompatibilities between fuel, liner, and clad were in evidence.

Intergranular tellurium cracking of nickel-based alloys in molten Li, Be, Th, U/F salt mixture

NASA Astrophysics Data System (ADS)

Ignatiev, Victor; Surenkov, Alexander; Gnidoy, Ivan; Kulakov, Alexander; Uglov, Vadim; Vasiliev, Alexander; Presniakov, Mikhail

2013-09-01

In Russia, R&D on Molten Salt Reactor (MSR) are concentrated now on fast/intermediate spectrum concepts which were recognized as long term alternative to solid fueled fast reactors due to their attractive features: strong negative feedback coefficients, easy in-service inspection, and simplified fuel cycle. For high-temperature MSR corrosion of the metallic container alloy in primary circuit is the primary concern. Key problem receiving current attention include surface fissures in Ni-based alloys probably arising from fission product tellurium attack. This paper summarizes results of corrosion tests conducted recently to study effect of oxidation state in selected fuel salt on tellurium attack and to develop means of controlling tellurium cracking in the special Ni-based alloys recently developed for molten salt actinide recycler and tranforming (MOSART) system. Tellurium corrosion of Ni-based alloys was tested at temperatures up to 750 °C in stressed and unloaded conditions in molten LiF-BeF2 salt mixture fueled by about 20 mol% of ThF4 and 2 mol% of UF4 at different [U(IV)]/[U(III)] ratios: 0.7, 4, 20, 100 and 500. Following Ni-based alloys (in mass%): HN80М-VI (Mo—12, Cr—7.6, Nb—1.5), HN80МТY (Mo—13, Cr—6.8, Al—1.1, Ti—0.9), HN80МТW (Mo—9.4, Cr—7.0, Ti—1.7, W—5.5) and ЕМ-721 (W—25.2, Cr—5.7, Ti—0.17) were used for the study in the corrosion facility. If the redox state the fuel salt is characterized by uranium ratio [U(IV)]/[U(III)] < 1 the alloys' specimens get a more negative stationary electrode potential than equilibrium electrode potentials of some uranium intermetallic compounds and alloys with nickel and molybdenum. This leads to spontaneous behavior of alloy formation processes on the specimens' surface and further diffusion of uranium deep into the metallic phase. As consequence of this films of intermetallic compounds and alloys of nickel, molybdenum, tungsten with uranium are formed on the alloys specimens' surface

Unrestrained swelling of uranium-nitride fuel irradiated at temperatures ranging from 1100 to 1400 K (1980 to 2520 R)

NASA Technical Reports Server (NTRS)

Rohal, R. G.; Tambling, T. N.

1973-01-01

Six fuel pins were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a stainless steel (type 304L) clad. The pins were irradiated for approximately 4000 hours to burnups of about 2.0 atom percent uranium. The average clad surface temperature during irradiation was about 1100 K (1980 deg R). Since stainless steel has a very low creep strength relative to that of UN at this temperature, these tests simulated unrestrained swelling of UN. The tests indicated that at 1 percent uranium atom burnup the unrestrained diametrical swelling of UN is about 0.5, 0.8, and 1.0 percent at 1223, 1264, and 1306 K (2200, deg 2273 deg, and 2350 deg R), respectively. The tests also indicated that the irradiation induced swelling of unrestrained UN fuel pellets appears to be isotropic.

Equilibrium of molybdenum in selected extraction systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tkac, Peter; Paulenova, Alena

2007-07-01

The concentration of molybdenum(VI) in dissolved irradiated nuclear fuel is comparable with the concentrations of Tc, Am and Np. Therefore it is of big interest to understand its behavior under conditions related to the UREX/TRUEX process. The effect of the poly-speciation of molybdenum in aqueous solution on its extraction by neutral solvents TBP and CMPO/TBP was studied. Extraction yields of molybdenum decreased significantly when AHA was added to aqueous phase. Our investigation confirmed a strong ability of the aceto-hydroxamic acid to form complexes with Mo in high acidic solutions. Spectroscopic data (UV-Vis) confirmed that a fraction of the Mo(VI)-AHA complexmore » is present in the organic phase after extraction. (authors)« less

Precipitation of α' in neutron irradiated commercial FeCrAl alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Littrell, Kenneth C.; Briggs, Samuel A.

2017-08-17

In this paper, Alkrothal 720 and Kanthal APMT™, two commercial FeCrAl alloys, were neutron irradiated up to damage doses of 7.0 displacements per atom (dpa) in the temperature range of 320 to 382 °C to characterize the α' precipitation in these alloys using small-angle neutron scattering. Both alloys exhibited α' precipitation. Kanthal APMT™ exhibited higher number densities and volume fraction, a result attributed to its higher Cr content compared with Alkrothal 720. Finally, trends observed as a function of damage dose (dpa) are consistent with literature trends for both FeCr and FeCrAl alloys

Effects of compositional complexity on the ion-irradiation induced swelling and hardening in Ni-containing equiatomic alloys

DOE PAGES

Jin, K.; Lu, C.; Wang, L. M.; ...

2016-04-14

The impact of compositional complexity on the ion-irradiation induced swelling and hardening is studied in Ni and six Ni-containing equiatomic alloys with face-centered cubic structure. The irradiation resistance at the temperature of 500 °C is improved by controlling the number and, especially, the type of alloying elements. Alloying with Fe and Mn has a stronger influence on swelling reduction than does alloying with Co and Cr. Lastly, the quinary alloy NiCoFeCrMn, with known excellent mechanical properties, has shown 40 times higher swelling tolerance than nickel.

Calculated mammographic spectra confirmed with attenuation curves for molybdenum, rhodium, and tungsten targets.

PubMed

Blough, M M; Waggener, R G; Payne, W H; Terry, J A

1998-09-01

A model for calculating mammographic spectra independent of measured data and fitting parameters is presented. This model is based on first principles. Spectra were calculated using various target and filter combinations such as molybdenum/molybdenum, molybdenum/rhodium, rhodium/rhodium, and tungsten/aluminum. Once the spectra were calculated, attenuation curves were calculated and compared to measured attenuation curves. The attenuation curves were calculated and measured using aluminum alloy 1100 or high purity aluminum filtration. Percent differences were computed between the measured and calculated attenuation curves resulting in an average of 5.21% difference for tungsten/aluminum, 2.26% for molybdenum/molybdenum, 3.35% for rhodium/rhodium, and 3.18% for molybdenum/rhodium. Calculated spectra were also compared to measured spectra from the Food and Drug Administration [Fewell and Shuping, Handbook of Mammographic X-ray Spectra (U.S. Government Printing Office, Washington, D.C., 1979)] and a comparison will also be presented.

Synthesis, thermal stability and the effects of ion irradiation in amorphous Si-O-C alloys

NASA Astrophysics Data System (ADS)

Colón Santana, Juan A.; Mora, Elena Echeverría; Price, Lloyd; Balerio, Robert; Shao, Lin; Nastasi, Michael

2015-05-01

Amorphous films of Si-O-C alloys were synthesized via sputtering deposition at room temperature. These alloys were characterized using grazing incidence diffraction, both as a function of temperature and irradiation dose. It was found that the material retained its amorphous structure, both at high temperatures (up to 1200 °C) and ion irradiation doses up to 1.0 dpa. The depth profile from photoemission spectroscopy provided evidence of the oxidation state of these alloys and their atomic composition. The studies suggest that Si-O-C alloys might belong to a group of radiation tolerant materials suitable for applications in reactor-like harsh environments.

Ductile metal alloys, method for making ductile metal alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cockeram, Brian V.

A ductile alloy is provided comprising molybdenum, chromium and aluminum, wherein the alloy has a ductile to brittle transition temperature of about 300 C after radiation exposure. The invention also provides a method for producing a ductile alloy, the method comprising purifying a base metal defining a lattice; and combining the base metal with chromium and aluminum, whereas the weight percent of chromium is sufficient to provide solute sites within the lattice for point defect annihilation.

Impact of neutron irradiation on mechanical performance of FeCrAl alloy laser-beam weldments

NASA Astrophysics Data System (ADS)

Gussev, M. N.; Cakmak, E.; Field, K. G.

2018-06-01

Oxidation-resistant iron-chromium-aluminum (FeCrAl) alloys demonstrate better performance in Loss-of-Coolant Accidents, compared with austenitic- and zirconium-based alloys. However, further deployment of FeCrAl-based materials requires detailed characterization of their performance under irradiation; moreover, since welding is one of the key operations in fabrication of light water reactor fuel cladding, FeCrAl alloy weldment performance and properties also should be determined prior to and after irradiation. Here, advanced C35M alloy (Fe-13%Cr-5%Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions were characterized after neutron irradiation in Oak Ridge National Laboratory's High Flux Isotope Reactor at 1.8-1.9 dpa in a temperature range of 195-559 °C. Specimen sets included as-received (AR) materials and specimens after controlled laser-beam welding. Tensile tests with digital image correlation (DIC), scanning electron microscopy-electron back scatter diffraction analysis, fractography, and x-ray tomography analysis were performed. DIC allowed for investigating local yield stress in the weldments, deformation hardening behavior, and plastic anisotropy. Both AR and welded material revealed a high degree of radiation-induced hardening for low-temperature irradiation; however, irradiation at high-temperatures (i.e., 559 °C) had little overall effect on the mechanical performance.

Single-Phase Concentrated Solid-Solution Alloys: Bridging Intrinsic Transport Properties and Irradiation Resistance

DOE PAGES

Jin, Ke; Bei, Hongbin

2018-04-30

Single-phase concentrated solid-solution alloys (SP-CSAs), including high entropy alloys (HEAs), are compositionally complex but structurally simple, and provide a playground of tailoring material properties through modifying their compositional complexity. The recent progress in understanding the compositional effects on the energy and mass transport properties in a series of face-centered-cubic SP-CSAs is the focus of this review. Relatively low electrical and thermal conductivities, as well as small separations between the interstitial and vacancy migration barriers have been generally observed, but largely depend on the alloying constituents. We further discuss the impact of such intrinsic transport properties on their irradiation response; themore » linkage to the delayed damage accumulation, slow defect aggregation, and suppressed irradiation induced swelling and segregation has been presented. We emphasize that the number of alloying elements may not be a critical factor on both transport properties and the defect behaviors under ion irradiations. Furthermore, the recent findings have stimulated novel concepts in the design of new radiation-tolerant materials, but further studies are demanded to enable predictive models that can quantitatively bridge the transport properties to the radiation damage.« less

Single-Phase Concentrated Solid-Solution Alloys: Bridging Intrinsic Transport Properties and Irradiation Resistance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jin, Ke; Bei, Hongbin

Single-phase concentrated solid-solution alloys (SP-CSAs), including high entropy alloys (HEAs), are compositionally complex but structurally simple, and provide a playground of tailoring material properties through modifying their compositional complexity. The recent progress in understanding the compositional effects on the energy and mass transport properties in a series of face-centered-cubic SP-CSAs is the focus of this review. Relatively low electrical and thermal conductivities, as well as small separations between the interstitial and vacancy migration barriers have been generally observed, but largely depend on the alloying constituents. We further discuss the impact of such intrinsic transport properties on their irradiation response; themore » linkage to the delayed damage accumulation, slow defect aggregation, and suppressed irradiation induced swelling and segregation has been presented. We emphasize that the number of alloying elements may not be a critical factor on both transport properties and the defect behaviors under ion irradiations. Furthermore, the recent findings have stimulated novel concepts in the design of new radiation-tolerant materials, but further studies are demanded to enable predictive models that can quantitatively bridge the transport properties to the radiation damage.« less

NICKEL-BASE ALLOY

DOEpatents

Inouye, H.; Manly, W.D.; Roche, T.K.

1960-01-19

A nickel-base alloy was developed which is particularly useful for the containment of molten fluoride salts in reactors. The alloy is resistant to both salt corrosion and oxidation and may be used at temperatures as high as 1800 deg F. Basically, the alloy consists of 15 to 22 wt.% molybdenum, a small amount of carbon, and 6 to 8 wt.% chromium, the balance being nickel. Up to 4 wt.% of tungsten, tantalum, vanadium, or niobium may be added to strengthen the alloy.

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

DOE PAGES

Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

2015-06-21

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3Si 2 has been optimized and high phase purity U 3Si 2 has been successfully produced. Results are presentedmore » from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3Si 2 has been optimized and high phase purity U 3Si 2 has been successfully produced. Results are presentedmore » from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

Precipitation behavior of AlxCoCrFeNi high entropy alloys under ion irradiation

NASA Astrophysics Data System (ADS)

Yang, Tengfei; Xia, Songqin; Liu, Shi; Wang, Chenxu; Liu, Shaoshuai; Fang, Yuan; Zhang, Yong; Xue, Jianming; Yan, Sha; Wang, Yugang

2016-08-01

Materials performance is central to the satisfactory operation of current and future nuclear energy systems due to the severe irradiation environment in reactors. Searching for structural materials with excellent irradiation tolerance is crucial for developing the next generation nuclear reactors. Here, we report the irradiation responses of a novel multi-component alloy system, high entropy alloy (HEA) AlxCoCrFeNi (x = 0.1, 0.75 and 1.5), focusing on their precipitation behavior. It is found that the single phase system, Al0.1CoCrFeNi, exhibits a great phase stability against ion irradiation. No precipitate is observed even at the highest fluence. In contrast, numerous coherent precipitates are present in both multi-phase HEAs. Based on the irradiation-induced/enhanced precipitation theory, the excellent structural stability against precipitation of Al0.1CoCrFeNi is attributed to the high configurational entropy and low atomic diffusion, which reduces the thermodynamic driving force and kinetically restrains the formation of precipitate, respectively. For the multiphase HEAs, the phase separations and formation of ordered phases reduce the system configurational entropy, resulting in the similar precipitation behavior with corresponding binary or ternary conventional alloys. This study demonstrates the structural stability of single-phase HEAs under irradiation and provides important implications for searching for HEAs with higher irradiation tolerance.

Dissolved molybdenum and uranium in the Three Rivers of eastern Tibet

NASA Astrophysics Data System (ADS)

Noh, H.; Huh, Y.

2006-12-01

Three large rivers - the Chang Jiang (Yangtze), Mekong (Lancang Jiang) and Salween (Nu Jiang) - originate in eastern Tibet and run in close parallel over 300 km near the eastern Himalayan syntaxis. They flow across suture zones and faults generated by the collision of India and Eurasia. Sixty-five water samples were collected in summer of 1999 to 2004 and nine in winter of 2002 to 2003. The complex geologic makeup of the Three Rivers region (TRR) results in widely varying major and trace element compositions of the dissolved load. Two redox-sensitive elements, molybdenum (Mo) and uranium (U) were analyzed by ICP-MS, as potential proxies for weathering of sedimentary organic carbon and resultant generation of atmospheric carbon dioxide. Additionally, Mo constitutes an essential co-enzyme for biology. Mo concentration ranges from 0.76 to 21.3 nmol/kg (average: 6.24 nmol/kg, average of global rivers: ~5 nmol/kg (Martin and Meybeck, 1979)), and U concentration varies from 2.86 to 10.7 nmol/kg (average: 3.12 nmol/kg, average of global rivers: ~1 nmol/kg (Palmer and Edmond, 1993)). The highest values of Mo and U are observed in the headwater tributary sample of the Chang Jiang, where evaporite dissolution is dominant. Statistical analyses show that Mo is closely correlated with U (r = 0.713, p < 0.01) indicating similar source of Mo and U to river waters in the TRR. Inverse correlation with Si/total anions ratio suggests that their sources are non-silicate minerals. The correlation with sulfate supports the use Mo and U as proxies for weathering of reduced organic-rich sediments (Mo and SO4: r = 0.383, p < 0.01; U and SO4: r = 0.508, p < 0.01). Among the parameters tested (basin area, elevation, relief, slope, T, precipitation, potential-evapotranspiration, normalized difference vegetation index (NDVI), population density), best positive correlation (r>0.5) is shown between U and basin elevation, and negative correlation is shown between U and temperature, precipitation

Evaluation of defect formation in helium irradiated Y2O3 doped W-Ti alloys by positron annihilation and nanoindentation

NASA Astrophysics Data System (ADS)

Richter, Asta; Anwand, Wolfgang; Chen, Chun-Liang; Böttger, Roman

2017-10-01

Helium implanted tungsten-titanium ODS alloys are investigated using positron annihilation spectroscopy and nanoindentation. Titanium reduces the brittleness of the tungsten alloy, which is manufactured by mechanical alloying. The addition of Y2O3 nanoparticles increases the mechanical properties at elevated temperature and enhances irradiation resistance. Helium ion implantation was applied to simulate irradiation effects on these materials. The irradiation was performed using a 500 kV He ion implanter at fluences around 5 × 1015 cm-2 for a series of samples both at room temperature and at 600 °C. The microstructure and mechanical properties of the pristine and irradiated W-Ti-ODS alloy are compared with respect to the titanium and Y2O3 content. Radiation damage is studied by positron annihilation spectroscopy analyzing the lifetime and the Doppler broadening. Three types of helium-vacancy defects were detected after helium irradiation in the W-Ti-ODS alloy: small defects with high helium-to-vacancy ratio (low S parameter) for room temperature irradiation, larger open volume defects with low helium-to-vacancy ratio (high S parameter) at the surface and He-vacancy complexes pinned at nanoparticles deeper in the material for implantation at 600 °C. Defect induced hardness was studied by nanoindentation. A drastic hardness increase is observed after He ion irradiation both for room temperature and elevated irradiation temperature of 600 °C. The Ti alloyed tungsten-ODS is more affected by the hardness increase after irradiation compared to the pure W-ODS alloy.

BRAZING ALLOYS

DOEpatents

Donnelly, R.G.; Gilliland, R.G.; Slaughter, G.M.

1963-02-26

A brazing alloy which, in the molten state, is characterized by excellent wettability and flowability, said alloy being capable of forming a corrosion resistant brazed joint wherein at least one component of said joint is graphite and the other component is a corrosion resistant refractory metal, said alloy consisting essentially of 20 to 50 per cent by weight of gold, 20 to 50 per cent by weight of nickel, and 15 to 45 per cent by weight of molybdenum. (AEC)

Effect of γ-IRRADIATION on the Mechanical Properties of Al-Cu Alloy

NASA Astrophysics Data System (ADS)

Abo-Elsoud, M.; Ismail, H.; Sobhy, Maged S.

SEM observations and Vickers hardness tests were performed to identify the irradiation effects. γ-irradiation effect during the aging hardening process can be explained depending on the composition of the alloy and is used to derive quantitative information on the kinetics of the transformation precipitates. Increasing the Cu content of an Al-Cu alloy can improve the aging hardness. The present results of the hardness behavior, with SEM observations of surveillance specimens at different doses, suggest that the radiation-induced defects are probably complex valence-solute clusters. These clusters act as nuclei for the precipitation of θ-Al2Cu type. This can be effectively utilized to study the systematics of nucleation of precipitates at vacancy-type defects. γ-irradiation probably plays the key role in defects responsible for material strengthening and embrittlement.

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.

1973-01-01

A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

Nanocluster irradiation evolution in Fe-9%Cr ODS and ferritic-martensitic alloys

NASA Astrophysics Data System (ADS)

Swenson, M. J.; Wharry, J. P.

2017-12-01

The objective of this study is to evaluate the influence of dose rate and cascade morphology on nanocluster evolution in a model Fe-9%Cr oxide dispersion strengthened steel and the commercial ferritic/martensitic (F/M) alloys HCM12A and HT9. We present a large, systematic data set spanning the three alloys, three irradiating particle types, four orders of magnitude in dose rate, and doses ranging 1-100 displacements per atom over 400-500 °C. Nanoclusters are characterized using atom probe tomography. ODS oxide nanoclusters experience partial dissolution after irradiation due to inverse Ostwald ripening, while F/M nanoclusters undergo Ostwald ripening. Damage cascade morphology is indicative of nanocluster number density evolution. Finally, the effects of dose rate on nanocluster morphology provide evidence for a temperature dilation theory, which purports that a negative temperature shift is necessary for higher dose rate irradiations to emulate nanocluster evolution in lower dose rate irradiations.

Dislocation loop formation in model FeCrAl alloys after neutron irradiation below 1 dpa

DOE PAGES

Field, Kevin G.; Briggs, Samuel A.; Sridharan, Kumar; ...

2017-08-01

FeCrAl alloys with varying compositions and microstructures are under consideration for accident-tolerant fuel cladding, but limited details exist on dislocation loop formation and growth for this class of alloys under neutron irradiation. Four model FeCrAl alloys with chromium contents ranging from 10.01 to 17.51 wt % and alunimum contents of 4.78 to 2.93 wt % were neutron irradiated to doses of 0.3–0.8 displacements per atom (dpa) at temperatures of 335–355°C. On-zone STEM imaging revealed a mixed population of black dots and larger dislocation loops with either a/2< 111 > or a< 100 > Burgers vectors. Weak composition dependencies were observedmore » and varied depending on whether the defect size, number density, or ratio of defect types was of interest. Here, the results were found to mirror those of previous studies on FeCrAl and FeCr alloys irradiated under similar conditions, although distinct differences exist.« less

Dislocation loop formation in model FeCrAl alloys after neutron irradiation below 1 dpa

NASA Astrophysics Data System (ADS)

Field, Kevin G.; Briggs, Samuel A.; Sridharan, Kumar; Yamamoto, Yukinori; Howard, Richard H.

2017-11-01

FeCrAl alloys with varying compositions and microstructures are under consideration for accident-tolerant fuel cladding, but limited details exist on dislocation loop formation and growth for this class of alloys under neutron irradiation. Four model FeCrAl alloys with chromium contents ranging from 10.01 to 17.51 wt % and aluminum contents of 4.78 to 2.93 wt % were neutron irradiated to doses of 0.3-0.8 displacements per atom (dpa) at temperatures of 335-355 °C. On-zone STEM imaging revealed a mixed population of black dots and larger dislocation loops with either a / 2 〈 111 〉 or a 〈 100 〉 Burgers vectors. Weak composition dependencies were observed and varied depending on whether the defect size, number density, or ratio of defect types was of interest. Results were found to mirror those of previous studies on FeCrAl and FeCr alloys irradiated under similar conditions, although distinct differences exist.

Laser irradiation effects on the surface, structural and mechanical properties of Al-Cu alloy 2024

NASA Astrophysics Data System (ADS)

Yousaf, Daniel; Bashir, Shazia; Akram, Mahreen; kalsoom, Umm-i.-; Ali, Nisar

2014-02-01

Laser irradiation effects on surface, structural and mechanical properties of Al-Cu-Mg alloy (Al-Cu alloy 2024) have been investigated. The specimens were irradiated for various fluences ranging from 3.8 to 5.5 J/cm2 using an Excimer (KrF) laser (248 nm, 18 ns, 30 Hz) under vacuum environment. The surface and structural modifications of the irradiated targets have been investigated by scanning electron microscope (SEM) and X-ray diffractometer (XRD), respectively. SEM analysis reveals the formation of micro-sized craters along the growth of periodic surface structures (ripples) at their peripheries. The size of the craters initially increases and then decreases by increasing the laser fluence. XRD analysis shows an anomalous trend in the peak intensity and crystallite size of the specimen irradiated for various fluences. A universal tensile testing machine and Vickers microhardness tester were employed in order to investigate the mechanical properties of the irradiated targets. The changes in yield strength, ultimate tensile strength and microhardness were found to be anomalous with increasing laser fluences. The changes in the surface and structural properties of Al-Cu alloy 2024 after laser irradiation have been associated with the changes in mechanical properties.

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

NASA Astrophysics Data System (ADS)

Aydogan, E.; Weaver, J. S.; Maloy, S. A.; El-Atwani, O.; Wang, Y. Q.; Mara, N. A.

2018-05-01

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al2O3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe2+ ion irradiation up to ∼16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two-beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size and a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α‧ precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ∼3.4 dpa and ∼16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aydogan, E.; Weaver, J. S.; Maloy, S. A.

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al 2O 3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe 2+ ion irradiation up to ~16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size andmore » a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α' precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ~3.4 dpa and ~16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.« less

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

DOE PAGES

Aydogan, E.; Weaver, J. S.; Maloy, S. A.; ...

2018-03-02

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al 2O 3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe 2+ ion irradiation up to ~16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size andmore » a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of <100> type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α' precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ~3.4 dpa and ~16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.« less

On the changing petroleum generation properties of Alum Shale over geological time caused by uranium irradiation

NASA Astrophysics Data System (ADS)

Yang, Shengyu; Schulz, Hans-Martin; Horsfield, Brian; Schovsbo, Niels H.; Noah, Mareike; Panova, Elena; Rothe, Heike; Hahne, Knut

2018-05-01

An interdisciplinary study was carried out to unravel organic-inorganic interactions caused by the radiogenic decay of uranium in the immature organic-rich Alum Shale (Middle Cambrian-Lower Ordovician). Based on pyrolysis experiments, uranium content is positively correlated with the gas-oil ratios and the aromaticities of both the free hydrocarbons residing in the rock and the pyrolysis products from its kerogen, indicating that irradiation has had a strong influence on organic matter composition overall and hence on petroleum potential. The Fourier Transform Ion Cyclotron Resonance mass spectrometry data reveal that macro-molecules in the uranium-rich Alum Shale samples are less alkylated than less irradiated counterparts, providing further evidence for structural alteration by α-particle bombardment. In addition, oxygen containing-compounds are enriched in the uranium-rich samples but are not easily degradable into low-molecular-weight products due to irradiation-induced crosslinking. Irradiation has induced changes in organic matter composition throughout the shale's entire ca. 500 Ma history, irrespective of thermal history. This factor has to be taken into account when reconstructing petroleum generation history. The Alum Shale's kerogen underwent catagenesis in the main petroleum kitchen area 420-340 Ma bp. Our calculations suggest the kerogen was much more aliphatic and oil-prone after deposition than that after extensive exposure to radiation. In addition, the gas sorption capacity of the organic matter in the Alum Shale can be assumed to have been less developed during Palaeozoic times, in contrast to results gained by sorption experiments performed at the present day, for the same reason. The kerogen reconstruction method developed here precludes overestimations of gas generation and gas retention in the Alum Shale by taking irradiation exposure into account and can thus significantly mitigate charge risk when applied in the explorations for both

TEM characterization of irradiated microstructure of Fe-9%Cr ODS and ferritic-martensitic alloys

NASA Astrophysics Data System (ADS)

Swenson, M. J.; Wharry, J. P.

2018-04-01

The objective of this study is to evaluate the effects of irradiation dose and dose rate on defect cluster (i.e. dislocation loops and voids) evolution in a model Fe-9%Cr oxide dispersion strengthened steel and commercial ferritic-martensitic steels HCM12A and HT9. Complimentary irradiations using Fe2+ ions, protons, or neutrons to doses ranging from 1 to 100 displacements per atom (dpa) at 500 °C are conducted on each alloy. The irradiated microstructures are characterized using transmission electron microscopy (TEM). Dislocation loops exhibit limited growth after 1 dpa upon Fe2+ and proton irradiation, while any voids observed are small and sparse. The average size and number density of loops are statistically invariant between Fe2+, proton, and neutron irradiated specimens at otherwise fixed irradiation conditions of ∼3 dpa, 500 °C. Therefore, we conclude that higher dose rate charged particle irradiations can reproduce the neutron irradiated loop microstructure with temperature shift governed by the invariance theory; this temperature shift is ∼0 °C for the high sink strength alloys studied herein.

Survey of Portions of the Chromium-Cobalt-Nickel-Molybdenum Quaternary System at 1,200 Degrees C

NASA Technical Reports Server (NTRS)

Rideout, Sheldon Paul; Beck, Paul A

1953-01-01

A survey was made of portions of the chromium-cobalt-nickel-molybdenum quaternary system at 1,200 degrees c by means of microscopic and x-ray diffraction studies. Since the face-centered cubic (alpha) solid solutions form the matrix of almost all practically useful high-temperature alloys, the solid solubility limits of the quaternary alpha phase were determined up to 20 percent molybdenum. The component cobalt-nickel-molybdenum, chromium-cobalt-molybdenum, and chromium-nickel-molybdenum ternary systems were also studied. The survey of these systems was confined to the determination of the boundaries of the face-centered cubic (alpha) solid solutions and of the phases coexisting with alpha at 1,200 degrees c.

Filler metal alloy for welding cast nickel aluminide alloys

DOEpatents

Santella, Michael L.; Sikka, Vinod K.

1998-01-01

A filler metal alloy used as a filler for welding east nickel aluminide alloys contains from about 15 to about 17 wt. % chromium, from about 4 to about 5 wt. % aluminum, equal to or less than about 1.5 wt. % molybdenum, from about 1 to about 4.5 wt. % zirconium, equal to or less than about 0.01 wt. % yttrium, equal to or less than about 0.01 wt. % boron and the balance nickel. The filler metal alloy is made by melting and casting techniques such as are melting the components of the filler metal alloy and east in copper chill molds.

Molybdenum, molybdenum oxides, and their electrochemistry.

PubMed

Saji, Viswanathan S; Lee, Chi-Woo

2012-07-01

The electrochemical behaviors of molybdenum and its oxides, both in bulk and thin film dimensions, are critical because of their widespread applications in steels, electrocatalysts, electrochromic materials, batteries, sensors, and solar cells. An important area of current interest is electrodeposited CIGS-based solar cells where a molybdenum/glass electrode forms the back contact. Surprisingly, the basic electrochemistry of molybdenum and its oxides has not been reviewed with due attention. In this Review, we assess the scattered information. The potential and pH dependent active, passive, and transpassive behaviors of molybdenum in aqueous media are explained. The major surface oxide species observed, reversible redox transitions of the surface oxides, pseudocapacitance and catalytic reduction are discussed along with carefully conducted experimental results on a typical molybdenum glass back contact employed in CIGS-based solar cells. The applications of molybdenum oxides and the electrodeposition of molybdenum are briefly reviewed. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

METHOD OF APPLYING NICKEL COATINGS ON URANIUM

DOEpatents

Gray, A.G.

1959-07-14

A method is presented for protectively coating uranium which comprises etching the uranium in an aqueous etching solution containing chloride ions, electroplating a coating of nickel on the etched uranium and heating the nickel plated uranium by immersion thereof in a molten bath composed of a material selected from the group consisting of sodium chloride, potassium chloride, lithium chloride, and mixtures thereof, maintained at a temperature of between 700 and 800 deg C, for a time sufficient to alloy the nickel and uranium and form an integral protective coating of corrosion-resistant uranium-nickel alloy.

Heterogeneous dislocation loop formation near grain boundaries in a neutron-irradiated commercial FeCrAl alloy

DOE PAGES

Field, Kevin G.; Briggs, Samuel A.; Hu, Xunxiang; ...

2016-11-01

FeCrAl alloys are an attractive materials class for nuclear power applications due to their increased environmental compatibility over more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300-400 °C have shown post-irradiation microstructures containing dislocation loops and Cr-rich ' phase. Although these initial works established the post-irradiation microstructures, little to no focus was applied towards the influence of pre-irradiation microstructures on this response. Here, a well annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 dpa at 382 °C and then the role of random high angle grain boundariesmore » on the spatial distribution and size of dislocation loops, dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and an increased size of dislocation loops in the vicinity directly adjacent to the grain boundary. Lastly, these results suggest the importance of the pre-irradiation microstructure on the radiation tolerance of FeCrAl alloys.« less

Amorphous metal alloy and composite

DOEpatents

Wang, Rong; Merz, Martin D.

1985-01-01

Amorphous metal alloys of the iron-chromium and nickel-chromium type have excellent corrosion resistance and high temperature stability and are suitable for use as a protective coating on less corrosion resistant substrates. The alloys are stabilized in the amorphous state by one or more elements of titanium, zirconium, hafnium, niobium, tantalum, molybdenum, and tungsten. The alloy is preferably prepared by sputter deposition.

PROCESSES FOR SEPARATING AND RECOVERING CONSTITUENTS OF NEUTRON IRRADIATED URANIUM

DOEpatents

Connick, R.E.; Gofman, J.W.; Pimentel, G.C.

1959-11-10

Processes are described for preparing plutonium, particularly processes of separating plutonium from uranium and fission products in neutron-irradiated uraniumcontaining matter. Specifically, plutonium solutions containing uranium, fission products and other impurities are contacted with reducing agents such as sulfur dioxide, uranous ion, hydroxyl ammonium chloride, hydrogen peroxide, and ferrous ion whereby the plutoninm is reduced to its fluoride-insoluble state. The reduced plutonium is then carried out of solution by precipitating niobic oxide therein. Uranium and certain fission products remain behind in the solution. Certain other fission products precipitate along with the plutonium. Subsequently, the plutonium and fission product precipitates are redissolved, and the solution is oxidized with oxidizing agents such as chlorine, peroxydisulfate ion in the presence of silver ion, permanganate ion, dichromate ion, ceric ion, and a bromate ion, whereby plutonium is oxidized to the fluoride-soluble state. The oxidized solution is once again treated with niobic oxide, thus precipitating the contamirant fission products along with the niobic oxide while the oxidized plutonium remains in solution. Plutonium is then recovered from the decontaminated solution.

The relationship between alloying elements and biologically produced ennoblement in natural waters.

PubMed

Eashwar, M; Lakshman Kumar, A; Hariharasuthan, R; Sreedhar, G

2015-01-01

A range of stainless steels, nickel-chromium and nickel-chromium-molybdenum alloys were exposed to coastal seawater from Mandapam (Indian Ocean) and freshwater from a perennial pond. Biofilms from both test waters produced an ennoblement of the open circuit potential (OCP) on all alloys as expected, which was slower but substantially larger in freshwater. In both waters an interesting relationship was perceived between the plateau OCP (Emax) and the mass percentage of the major alloying elements. In particular, iron exhibited strong positive correlations with Emax (r(2) ≥ 0.77; p < 0.0005), while the sum of chromium, nickel and molybdenum presented significant negative correlations (r(2) ≤ -0.81; p = 0.0002). Consistent with the regression analyses, Euclidean distance clustering yielded patterns where Inconel-600 and the nickel-chromium-molybdenum alloys had the smallest similarities of OCP with other alloys. The results emphatically reinforce a key role for surface passive films in the ennoblement phenomenon in natural waters.

Iron-based amorphous alloys and methods of synthesizing iron-based amorphous alloys

DOEpatents

Saw, Cheng Kiong; Bauer, William A.; Choi, Jor-Shan; Day, Dan; Farmer, Joseph C.

2016-05-03

A method according to one embodiment includes combining an amorphous iron-based alloy and at least one metal selected from a group consisting of molybdenum, chromium, tungsten, boron, gadolinium, nickel phosphorous, yttrium, and alloys thereof to form a mixture, wherein the at least one metal is present in the mixture from about 5 atomic percent (at %) to about 55 at %; and ball milling the mixture at least until an amorphous alloy of the iron-based alloy and the at least one metal is formed. Several amorphous iron-based metal alloys are also presented, including corrosion-resistant amorphous iron-based metal alloys and radiation-shielding amorphous iron-based metal alloys.

Irradiation effects in oxide dispersion strengthened (ODS) Ni-base alloys for Gen. IV nuclear reactors

NASA Astrophysics Data System (ADS)

Oono, Naoko; Ukai, Shigeharu; Kondo, Sosuke; Hashitomi, Okinobu; Kimura, Akihiko

2015-10-01

Oxide particle dispersion strengthened (ODS) Ni-base alloys are irradiated by using simulation technique (Fe/He dual-ion irradiation) to investigate the reliability to Gen. IV high-temperature reactors. The fine oxide particles with less than 10 nm in average size and approximately 8.0 × 1022 m-3 in number density remained after 101 dpa irradiation. The tiny helium bubbles were inside grains, not at grain-boundaries; it is advantageous effect of oxide particles which trap the helium atoms at the particle-matrix interface. Ni-base ODS alloys demonstrated their great ability to overcome He embrittlement.

Mechanical properties and microstructural change of W–Y2O3 alloy under helium irradiation

PubMed Central

Tan, Xiaoyue; Luo, Laima; Chen, Hongyu; Zhu, Xiaoyong; Zan, Xiang; Luo, Guangnan; Chen, Junling; Li, Ping; Cheng, Jigui; Liu, Dongping; Wu, Yucheng

2015-01-01

A wet-chemical method combined with spark plasma sintering was used to prepare a W–Y2O3 alloy. High-temperature tensile tests and nano-indentation microhardness tests were used to characterize the mechanical properties of the alloy. After He-ion irradiation, fuzz and He bubbles were observed on the irradiated surface. The irradiation embrittlement was reflected by the crack indentations formed during the microhardness tests. A phase transformation from α-W to γ-W was investigated by X-ray diffraction (XRD) and transmission electron microscopy (TEM). Polycrystallization and amorphization were also observed in the irradiation damage layer. The W materials tended to exhibit lattice distortion, amorphization, polycrystallization and phase transformation under He-ion irradiation. The transformation mechanism predicted by the atomic lattice model was consistent with the available experimental observations. These findings clarify the mechanism of the structural transition of W under ion irradiation and provide a clue for identifying materials with greater irradiation resistance. PMID:26227480

Liquid uranium alloy-helium fission reactor

DOEpatents

Minkov, V.

1984-06-13

This invention describes a nuclear fission reactor which has a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200 to 1800/sup 0/C range, and even higher to 2500/sup 0/C.

Oxidation resistant alloys, method for producing oxidation resistant alloys

DOEpatents

Dunning, John S.; Alman, David E.

2002-11-05

A method for producing oxidation-resistant austenitic alloys for use at temperatures below 800 C. comprising of: providing an alloy comprising, by weight %: 14-18% chromium, 15-18% nickel, 1-3% manganese, 1-2% molybdenum, 2-4% silicon, 0% aluminum and the balance being iron; heating the alloy to 800 C. for between 175-250 hours prior to use in order to form a continuous silicon oxide film and another oxide film. The method provides a means of producing stainless steels with superior oxidation resistance at temperatures above 700 C. at a low cost

Ablation experiment and threshold calculation of titanium alloy irradiated by ultra-fast pulse laser

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zheng, Buxiang; Jiang, Gedong; Wang, Wenjun, E-mail: wenjunwang@mail.xjtu.edu.cn

The interaction between an ultra-fast pulse laser and a material's surface has become a research hotspot in recent years. Micromachining of titanium alloy with an ultra-fast pulse laser is a very important research direction, and it has very important theoretical significance and application value in investigating the ablation threshold of titanium alloy irradiated by ultra-fast pulse lasers. Irradiated by a picosecond pulse laser with wavelengths of 1064 nm and 532 nm, the surface morphology and feature sizes, including ablation crater width (i.e. diameter), ablation depth, ablation area, ablation volume, single pulse ablation rate, and so forth, of the titanium alloymore » were studied, and their ablation distributions were obtained. The experimental results show that titanium alloy irradiated by a picosecond pulse infrared laser with a 1064 nm wavelength has better ablation morphology than that of the green picosecond pulse laser with a 532 nm wavelength. The feature sizes are approximately linearly dependent on the laser pulse energy density at low energy density and the monotonic increase in laser pulse energy density. With the increase in energy density, the ablation feature sizes are increased. The rate of increase in the feature sizes slows down gradually once the energy density reaches a certain value, and gradually saturated trends occur at a relatively high energy density. Based on the linear relation between the laser pulse energy density and the crater area of the titanium alloy surface, and the Gaussian distribution of the laser intensity on the cross section, the ablation threshold of titanium alloy irradiated by an ultra-fast pulse laser was calculated to be about 0.109 J/cm{sup 2}.« less

Irradiation of organic matter by uranium decay in the Alum Shale, Sweden

NASA Astrophysics Data System (ADS)

Lewan, M. D.; Buchardt, B.

1989-06-01

The Alum Shale of Sweden contains black shales with anomalously high uranium concentrations in excess of 100 ppm. Syngenetic or early diagenetic origin of this uranium indicates that organic matter within these shales has been irradiated by decaying uranium for approximately 500 Ma. Radiation-induced polymerization of alkanes through a free-radical cross-linking mechanism appears to be responsible for major alterations within the irradiated organic matter. Specific radiation-induced alterations include generation of condensate-like oils at reduced yields from hydrous pyrolysis experiments, decrease in atomic H/C ratios of kerogens, decrease in bitumen/organic-carbon ratios, and a relative increase in low-molecular weight triaromatic steroid hydrocarbons. Conversely, stable carbon isotopes of kerogens, reflectance of vitrinite-like macerais, oil-generation kinetics, and isomerization of 20R to 20S αα C 29-steranes were not affected by radiation. The radiation dosage needed to cause the alterations observed in the Alum Shale has been estimated to be in excess of 10 5 Mrads with respect to organic carbon. This value is used to estimate the potential for radiation damage to thermally immature organic matter in black shales through the geological rock record. High potential for radiation damage is not likely in Cenozoic and Mesozoic black shales but becomes more likely in lower Paleozoic and Precambrian black shales.

Dosimetry characterization of the Godiva Reactor under burst conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hickman, D. P.; Heinrichs, D. P.; Hudson, R.

2017-06-22

A series of sixteen (16) burst irradiations were performed in May 2014, fifteen of which were part of an international collaboration to characterize the Godiva IV fast burst reactor at the National Criticality Experiments Research Center (NCERC). Godiva IV is a bare cylindrical assembly of approximately 65 kg of highly enriched uranium fuel (93.2% 235U metal alloyed with 1.5% molybdenum for strength) and is designed to perform controlled prompt critical excursions (Myers 2010, Goda 2013). Twelve of the irradiations were dedicated to neutron spectral measurements using a Bonner multiple sphere spectrometer. Three irradiations, with core temperature increases of 71.1°C, 136.9°C,more » and 229.9°C, were performed for generating comparative fluence data, establishing corrections for varying heights, testing linearity with burst temperature, and establishing gamma dose characteristics.« less

SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS FROM NEUTRON- BOMBARDED URANIUM

DOEpatents

Martin, A.E.; Johnson, I.; Burris, L. Jr.; Winsch, I.O.; Feder, H.M.

1962-11-13

A process is given for removing plutonium and/or fission products from uranium fuel. The fuel is dissolved in molten zinc--magnesium (10 to 18% Mg) alloy, more magnesium is added to obtain eutectic composition whereby uranium precipitates, and the uranium are separated from the Plutoniumand fission-product- containing eutectic. (AEC)

Effects of Fe concentration on the ion-irradiation induced defect evolution and hardening in Ni-Fe solid solution alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jin, Ke; Guo, Wei; Lu, Chenyang

Understanding alloying effects on the irradiation response of structural materials is pivotal in nuclear engineering. In order to systematically explore the effects of Fe concentration on the irradiation-induced defect evolution and hardening in face-centered cubic Ni-Fe binary solid solution alloys, single crystalline Ni-xFe (x = 0–60 at%) alloys have been grown and irradiated with 1.5 MeV Ni ions. The irradiations have been performed over a wide range of fluences from 3 × 10 13 to 3 × 10 16 cm -2 at room temperature. Ion channeling technique has shown reduced damage accumulation with increasing Fe concentration in the low fluencemore » regime, which is consistent to the results from molecular dynamic simulations. We did not observe any irradiation-induced compositional segregation in atom probe tomography within the detection limit, even in the samples irradiated with high fluence Ni ions. Transmission electron microscopy analyses have further demonstrated that the defect size significantly decreases with increasing Fe concentration, indicating a delay in defect evolution. Furthermore, irradiation induced hardening has been measured by nanoindentation tests. Ni and the Ni-Fe alloys have largely different initial hardness, but they all follow a similar trend for the increase of hardness as a function of irradiation fluence.« less

Effects of Fe concentration on the ion-irradiation induced defect evolution and hardening in Ni-Fe solid solution alloys

DOE PAGES

Jin, Ke; Guo, Wei; Lu, Chenyang; ...

2016-12-01

Understanding alloying effects on the irradiation response of structural materials is pivotal in nuclear engineering. In order to systematically explore the effects of Fe concentration on the irradiation-induced defect evolution and hardening in face-centered cubic Ni-Fe binary solid solution alloys, single crystalline Ni-xFe (x = 0–60 at%) alloys have been grown and irradiated with 1.5 MeV Ni ions. The irradiations have been performed over a wide range of fluences from 3 × 10 13 to 3 × 10 16 cm -2 at room temperature. Ion channeling technique has shown reduced damage accumulation with increasing Fe concentration in the low fluencemore » regime, which is consistent to the results from molecular dynamic simulations. We did not observe any irradiation-induced compositional segregation in atom probe tomography within the detection limit, even in the samples irradiated with high fluence Ni ions. Transmission electron microscopy analyses have further demonstrated that the defect size significantly decreases with increasing Fe concentration, indicating a delay in defect evolution. Furthermore, irradiation induced hardening has been measured by nanoindentation tests. Ni and the Ni-Fe alloys have largely different initial hardness, but they all follow a similar trend for the increase of hardness as a function of irradiation fluence.« less

Local structure of NiPd solid solution alloys and its response to ion irradiation

DOE PAGES

Zhang, Fuxiang; Ullah, Mohammad Wali; Zhao, Shijun; ...

2018-04-27

The local structure of Ni$-$Pd solid solution alloys with compositions of Ni 80Pd 20 and Ni 50Pd 50 was investigated with anomalous X-ray diffraction, X-ray absorption and theoretical calculation/simulation. The fcc lattice is distorted for both alloys, and the Pd$-$Pd atomic pair distance is +4.4% and +1.4% larger than ideal values in Ni 80Pd 20 and Ni 50Pd 50 alloys, respectively. The corresponding atomic pair distance of Ni$-$Ni is -1.8% and -3.0% less than the ideal values. Different short-range orders in the alloys were quantitatively identified at the atomic level. In Ni 80Pd 20, Pd atoms are likely to formmore » Pd$-$Pd pairs, while Pd atoms are connected with Pd atoms in the second shell in the equiatomic solid solution alloy. Upon ion irradiation, little change of interatomic distance, but modification of chemical short-range order was observed. The number of Pd$-$Pd pairs decreases to the lowest value at 0.1 dpa, and further irradiation make it increase.« less

Local structure of NiPd solid solution alloys and its response to ion irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Fuxiang; Ullah, Mohammad Wali; Zhao, Shijun

The local structure of Ni$-$Pd solid solution alloys with compositions of Ni 80Pd 20 and Ni 50Pd 50 was investigated with anomalous X-ray diffraction, X-ray absorption and theoretical calculation/simulation. The fcc lattice is distorted for both alloys, and the Pd$-$Pd atomic pair distance is +4.4% and +1.4% larger than ideal values in Ni 80Pd 20 and Ni 50Pd 50 alloys, respectively. The corresponding atomic pair distance of Ni$-$Ni is -1.8% and -3.0% less than the ideal values. Different short-range orders in the alloys were quantitatively identified at the atomic level. In Ni 80Pd 20, Pd atoms are likely to formmore » Pd$-$Pd pairs, while Pd atoms are connected with Pd atoms in the second shell in the equiatomic solid solution alloy. Upon ion irradiation, little change of interatomic distance, but modification of chemical short-range order was observed. The number of Pd$-$Pd pairs decreases to the lowest value at 0.1 dpa, and further irradiation make it increase.« less

Shear Punch Testing on ATR Irradiated MA956 FeCrAl Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saleh, Tarik A.; Quintana, Matthew Estevan; Romero, Tobias J.

2017-06-13

The shear punch testing of irradiated and control MA956 (FeCrAl) Alloy from the NSUF-ATR-UCSB irradiation is presented. This is the first data taken on a new shear punch fixture design to test three 1.5mm punches from each 8mm x 0.5mm Disc Multipurpose Coupon (DMC). Samples were irradiated to 6.1dpa at a temperature of 315°C and 6.2 dpa at 400°C.

Filler metal alloy for welding cast nickel aluminide alloys

DOEpatents

Santella, M.L.; Sikka, V.K.

1998-03-10

A filler metal alloy used as a filler for welding cast nickel aluminide alloys contains from about 15 to about 17 wt. % chromium, from about 4 to about 5 wt. % aluminum, equal to or less than about 1.5 wt. % molybdenum, from about 1 to about 4.5 wt. % zirconium, equal to or less than about 0.01 wt. % yttrium, equal to or less than about 0.01 wt. % boron and the balance nickel. The filler metal alloy is made by melting and casting techniques such as are melting the components of the filler metal alloy and cast in copper chill molds. 3 figs.

Heterogeneous dislocation loop formation near grain boundaries in a neutron-irradiated commercial FeCrAl alloy

NASA Astrophysics Data System (ADS)

Field, Kevin G.; Briggs, Samuel A.; Hu, Xunxiang; Yamamoto, Yukinori; Howard, Richard H.; Sridharan, Kumar

2017-01-01

FeCrAl alloys are an attractive class of materials for nuclear power applications because of their increased environmental compatibility compared with more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300 and 400 °C have shown post-irradiation microstructures containing dislocation loops and a Cr-rich α‧ phase. Although these initial studies established the post-irradiation microstructures, there was little to no focus on understanding the influence of pre-irradiation microstructures on this response. In this study, a well-annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 displacements per atom (dpa) at 382 °C and then the effect of random high-angle grain boundaries on the spatial distribution and size of a〈100〉 dislocation loops, a/2〈111〉 dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with a/2〈111〉 dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and a〈100〉 dislocation loops exhibiting an increased size in the vicinity of the grain boundary. These results suggest the importance of the pre-irradiation microstructure and, specifically, defect sink density spacing to the radiation tolerance of FeCrAl alloys.

ELECTROLYSIS OF THORIUM AND URANIUM

DOEpatents

Hansen, W.N.

1960-09-01

An electrolytic method is given for obtaining pure thorium, uranium, and thorium-uranium alloys. The electrolytic cell comprises a cathode composed of a metal selected from the class consisting of zinc, cadmium, tin, lead, antimony, and bismuth, an anode composed of at least one of the metals selected from the group consisting of thorium and uranium in an impure state, and an electrolyte composed of a fused salt containing at least one of the salts of the metals selected from the class consisting of thorium, uranium. zinc, cadmium, tin, lead, antimony, and bismuth. Electrolysis of the fused salt while the cathode is maintained in the molten condition deposits thorium, uranium, or thorium-uranium alloys in pure form in the molten cathode which thereafter may be separated from the molten cathode product by distillation.

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Magoulas, V. E.

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium,more » and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.« less

Further Charpy impact test results of low activation ferritic alloys, irradiated at 430{degrees}C to 67 dpa

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

1997-04-01

Miniature CVN specimens of four ferritic alloys, GA3X, F82H, GA4X and HT9, have been impact tested following irradiation at 430{degrees}C to 67 dpa. Comparison of the results with those of the previously tested lower dose irradiation condition indicates that the GA3X and F82H alloys, two primary candidate low activation alloys, exhibit virtually identical behavior following irradiation at 430{degrees}C to {approximately}67 dpa and at 370{degrees}C to {approximately}15 dpa. Very little shift is observed in either DBTT or USE relative to the unirradiated condition. The shifts in DBTT and USE observed in both GA4X and HT9 were smaller after irradiation at 430{degrees}Cmore » to {approximately}67 dpa than after irradiation at 370{degrees}C to {approximately}15 dpa.« less

Microstructure of RERTR DU-alloys irradiated with krypton ions up to 100 dpa

NASA Astrophysics Data System (ADS)

Gan, J.; Keiser, D. D., Jr.; Miller, B. D.; Wachs, D. M.; Allen, T. R.; Kirk, M.; Rest, J.

2011-04-01

The radiation stability of the interaction product formed at the fuel-matrix interface of research reactor dispersion fuels, under fission-product bombardment, has a strong impact on fuel performance. Three depleted uranium alloys were cast that consisted of the following five phases to be investigated: U(Si, Al) 3, (U, Mo)(Si, Al) 3, UMo 2Al 20, U 6Mo 4Al 43, and UAl 4. Irradiation of transmission electron microscopy (TEM) disc samples with 500-keV Kr ions at 200 °C to doses up to ˜100 displacements per atom (dpa) were conducted using a 300-keV electron microscope equipped with an ion accelerator. TEM results show that the U(Si, Al) 3 and UAl 4 phases remain crystalline at 100 dpa without forming voids. The (U, Mo)(Si, Al) 3 and UMo 2Al 20 phases become amorphous at 1 and ˜2 dpa, respectively, and show no evidence of voids at 100 dpa. The U 6Mo 4Al 43 phase goes to amorphous at less than 1 dpa and reveals high density voids at 100 dpa.

Accelerated Irradiations for High Dose Microstructures in Fast Reactor Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiao, Zhijie

The objective of this project is to determine the extent to which high dose rate, self-ion irradiation can be used as an accelerated irradiation tool to understand microstructure evolution at high doses and temperatures relevant to advanced fast reactors. We will accomplish the goal by evaluating phase stability and swelling of F-M alloys relevant to SFR systems at very high dose by combining experiment and modeling in an effort to obtain a quantitative description of the processes at high and low damage rates.

Atom redistribution and multilayer structure in NiTi shape memory alloy induced by high energy proton irradiation

NASA Astrophysics Data System (ADS)

Wang, Haizhen; Yi, Xiaoyang; Zhu, Yingying; Yin, Yongkui; Gao, Yuan; Cai, Wei; Gao, Zhiyong

2017-10-01

The element distribution and surface microstructure in NiTi shape memory alloys exposed to 3 MeV proton irradiation were investigated. Redistribution of the alloying element and a clearly visible multilayer structure consisting of three layers were observed on the surface of NiTi shape memory alloys after proton irradiation. The outermost layer consists primarily of a columnar-like TiH2 phase with a tetragonal structure, and the internal layer is primarily comprised of a bcc austenite phase. In addition, the Ti2Ni phase, with an fcc structure, serves as the transition layer between the outermost and internal layer. The above-mentioned phenomenon is attributed to the preferential sputtering of high energy protons and segregation induced by irradiation.

PURIFICATION OF URANIUM FUELS

DOEpatents

Niedrach, L.W.; Glamm, A.C.

1959-09-01

An electrolytic process of refining or decontaminating uranium is presented. The impure uranium is made the anode of an electrolytic cell. The molten salt electrolyte of this cell comprises a uranium halide such as UF/sub 4/ or UCl/sub 3/ and an alkaline earth metal halide such as CaCl/sub 2/, BaF/sub 2/, or BaCl/sub 2/. The cathode of the cell is a metal such as Mn, Cr, Co, Fe, or Ni which forms a low melting eutectic with U. The cell is operated at a temperature below the melting point of U. In operation the electrodeposited uranium becomes alloyed with the metal of the cathode, and the low melting alloy thus formed drips from the cathode.

Evaluation of irradiation effects of near-infrared free-electron-laser of silver alloy for dental application.

PubMed

Kuwada-Kusunose, Takao; Kusunose, Alisa; Wakami, Masanobu; Takebayashi, Chikako; Goto, Haruhiko; Aida, Masahiro; Sakai, Takeshi; Nakao, Keisuke; Nogami, Kyoko; Inagaki, Manabu; Hayakawa, Ken; Suzuki, Kunihiro; Sakae, Toshiro

2017-08-01

In the application of lasers in dentistry, there is a delicate balance between the benefits gained from laser treatment and the heat-related damage arising from laser irradiation. Hence, it is necessary to understand the different processes associated with the irradiation of lasers on dental materials. To obtain insight for the development of a safe and general-purpose laser for dentistry, the present study examines the physical effects associated with the irradiation of a near-infrared free-electron laser (FEL) on the surface of a commonly used silver dental alloy. The irradiation experiments using a 2900-nm FEL confirmed the formation of a pit in the dental alloy. The pit was formed with one macro-pulse of FEL irradiation, therefore, suggesting the possibility of efficient material processing with an FEL. Additionally, there was only a slight increase in the silver alloy temperature (less than 0.9 °C) despite the long duration of FEL irradiation, thus inferring that fixed prostheses in the oral cavity can be processed by FEL without thermal damage to the surrounding tissue. These results indicate that dental hard tissues and dental materials in the oral cavity can be safely and efficiently processed by the irradiation of a laser, which has the high repetition rate of a femtosecond laser pulse with a wavelength around 2900 nm.

PLUTONIUM-URANIUM ALLOY

DOEpatents

Coffinberry, A.S.; Schonfeld, F.W.

1959-09-01

Pu-U-Fe and Pu-U-Co alloys suitable for use as fuel elements tn fast breeder reactors are described. The advantages of these alloys are ease of fabrication without microcracks, good corrosion restatance, and good resistance to radiation damage. These advantages are secured by limitation of the zeta phase of plutonium in favor of a tetragonal crystal structure of the U/sub 6/Mn type.

Oxidation resistant alloys, method for producing oxidation resistant alloys

DOEpatents

Dunning, John S.; Alman, David E.

2002-11-05

A method for producing oxidation-resistant austenitic alloys for use at temperatures below 800.degree. C. comprising of: providing an alloy comprising, by weight %: 14-18% chromium, 15-18% nickel, 1-3% manganese, 1-2% molybdenum, 2-4% silicon, 0% aluminum and the balance being iron; heating the alloy to 800.degree. C. for between 175-250 hours prior to use in order to form a continuous silicon oxide film and another oxide film. The method provides a means of producing stainless steels with superior oxidation resistance at temperatures above 700.degree. C. at a low cost

Finite-element model to predict roll-separation force and defects during rolling of U-10Mo alloys

NASA Astrophysics Data System (ADS)

Soulami, Ayoub; Burkes, Douglas E.; Joshi, Vineet V.; Lavender, Curt A.; Paxton, Dean

2017-10-01

A major goal of the Convert Program of the U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA) is to enable high-performance research reactors to operate with low-enriched uranium rather than the high-enriched uranium currently used. To this end, uranium alloyed with 10 wt% molybdenum (U-10Mo) represents an ideal candidate because of its stable gamma phase, low neutron caption cross section, acceptable swelling response, and predictable irradiation behavior. However, because of the complexities of the fuel design and the need for rolled monolithic U-10Mo foils, new developments in processing and fabrication are necessary. This study used a finite-element code, LS-DYNA, as a predictive tool to optimize the rolling process. Simulations of the hot rolling of U-10Mo coupons encapsulated in low-carbon steel were conducted following two different schedules. Model predictions of the roll-separation force and roll pack thicknesses at different stages of the rolling process were compared with experimental measurements. The study reported here discussed various attributes of the rolled coupons revealed by the model (e.g., waviness and thickness non-uniformity like dog-boning). To investigate the influence of the cladding material on these rolling defects, other cases were simulated: hot rolling with alternative can materials, namely, 304 stainless steel and Zircaloy-2, and bare-rolling. Simulation results demonstrated that reducing the mismatch in strength between the coupon and can material improves the quality of the rolled sheet. Bare-rolling simulation results showed a defect-free rolled coupon. The finite-element model developed and presented in this study can be used to conduct parametric studies of several process parameters (e.g., rolling speed, roll diameter, can material, and reduction).

A Nuclear Reactor and Chemical Processing Design for Production of Molybdenum-99 with Crystalline Uranyl Nitrate Hexahydrate Fuel

NASA Astrophysics Data System (ADS)

Stange, Gary Michael

Medical radioisotopes are used in tens of millions of procedures every year to detect and image a wide variety of maladies and conditions in the human body. The most widely-used diagnostic radioisotope is technetium-99m, a metastable isomer of technetium-99 that is generated by the radioactive decay of molybdenum-99. For a number of reasons, the supply of molybdenum-99 has become unreliable and the techniques used to produce it have become unattractive. This has spurred the investigation of new technologies that avoid the use of highly enriched uranium to produce molybdenum-99 in the United States, where approximately half of the demand originates. The first goal of this research is to develop a critical nuclear reactor design powered by solid, discrete pins of low enriched uranium. Analyses of single-pin heat transfer and whole-core neutronics are performed to determine the required specifications. Molybdenum-99 is produced directly in the fuel of this reactor and then extracted through a series of chemical processing steps. After this extraction, the fuel is left in an aqueous state. The second goal of this research is to describe a process by which the uranium may be recovered from this spent fuel solution and reconstituted into the original fuel form. Fuel recovery is achieved through a crystallization step that generates solid uranyl nitrate hexahydrate while leaving the majority of fission products and transuranic isotopes in solution. This report provides background information on molybdenum-99 production and crystallization chemistry. The previously unknown thermal conductivity of the fuel material is measured. Following this is a description of the modeling and calculations used to develop a reactor concept. The operational characteristics of the reactor core model are analyzed and reported. Uranyl nitrate crystallization experiments have also been conducted, and the results of this work are presented here. Finally, a process flow scheme for uranium

METHOD FOR PURIFYING URANIUM

DOEpatents

Knighton, J.B.; Feder, H.M.

1960-04-26

A process is given for purifying a uranium-base nuclear material. The nuclear material is dissolved in zinc or a zinc-magnesium alloy and the concentration of magnesium is increased until uranium precipitates.

NICKEL COATED URANIUM ARTICLE

DOEpatents

Gray, A.G.

1958-10-01

Nickel coatings on uranium and various methods of obtaining such coatings are described. Specifically disclosed are such nickel or nickel alloy layers as barriers between uranium and aluminum- silicon, chromium, or copper coatings.

Evolution of irradiation-induced strain in an equiatomic NiFe alloy

DOE PAGES

Ullah, Mohammad W.; Zhang, Yanwen; Sellami, Neila; ...

2017-07-10

Here, we investigate the formation and accumulation of irradiation-induced atomic strain in an equiatomic NiFe concentrated solid-solution alloy using both atomistic simulations and x-ray diffraction (XRD) analysis of irradiated samples. Experimentally, the irradiations are performed using 1.5 MeV Ni ions to fluences ranging from 1 × 10 13 to 1 × 10 14 cm -2. The irradiation simulations are carried out by overlapping 5 keV Ni recoils cascades up to a total of 300 recoils. An increase of volumetric strain is observed at low dose, which is associated with production of point defects and small clusters. A relaxation of strainmore » occurs at higher doses, when large defect clusters, like dislocation loops, dominate.« less

Surface and structure modification induced by high energy and highly charged uranium ion irradiation in monocrystal spinel

NASA Astrophysics Data System (ADS)

Yang, Yitao; Zhang, Chonghong; Song, Yin; Gou, Jie; Zhang, Liqing; Meng, Yancheng; Zhang, Hengqing; Ma, Yizhun

2014-05-01

Due to its high temperature properties and relatively good behavior under irradiation, magnesium aluminate spinel (MgAl2O4) is considered as a possible material to be used as inert matrix for the minor actinides burning. In this case, irradiation damage is an unavoidable problem. In this study, high energy and highly charged uranium ions (290 MeV U32+) were used to irradiate monocrystal spinel to the fluence of 1.0 × 1013 ions/cm2 to study the modification of surface and structure. Highly charged ions carry large potential energy, when they interact with a surface, the release of potential energy results in the modification of surface. Atomic force microscopy (AFM) results showed the occurrence of etching on surface after uranium ion irradiation. The etching depth reached 540 nm. The surprising efficiency of etching is considered to be induced by the deposition of potential energy with high density. The X-ray diffraction results showed that the (4 4 0) diffraction peak obviously broadened after irradiation, which indicated that the distortion of lattice has occurred. After multi-peak Gaussian fitting, four Gaussian peaks were separated, which implied that a structure with different damage layers could be formed after irradiation.

Tailoring molybdenum nanostructure evolution by low-energy He+ ion irradiation

NASA Astrophysics Data System (ADS)

Tripathi, J. K.; Novakowski, T. J.; Hassanein, A.

2015-10-01

Mirror-finished polished molybdenum (Mo) samples were irradiated with 100 eV He+ ions as a function of ion fluence (using a constant flux of 7.2 × 1020 ions m-2 s-1) at normal incidence and at 923 K. Mo surface deterioration and nanoscopic fiber-form filament ("Mo fuzz") growth evolution were monitored by using field emission (FE) scanning electron (SEM) and atomic force (AFM) microscopy studies. Those studies confirm a reasonably clean and flat surface, up to several micrometer scales along with a few mechanical-polishing-induced scratches. However, He+ ion irradiation deteriorates the surface significantly even at 2.1 × 1023 ions m-2 fluence (about 5 min. irradiation time) and leads to evolution of homogeneously populated ∼75-nm-long Mo nanograins having ∼8 nm intergrain width. The primary stages of Mo fuzz growth, i.e., elongated half-cylindrical ∼70 nm nanoplatelets, and encapsulated bubbles of 20-45 nm in diameter and preferably within the grain boundaries of sub-micron-sized grains, were observed after 1.3 × 1024 ions m-2 fluence irradiation. Additionally, a sequential enhancement in the sharpness, density, and protrusions of Mo fuzz at the surface with ion fluence was also observed. Fluence- and flux-dependent studies have also been performed at 1223 K target temperature (beyond the temperature window for Mo fuzz formation). At a constant fluence of 2.6 × 1024 ions m-2, 7.2 × 1020 ions m-2 s-1 flux generates a homogeneous layered and stacked nanodiscs of ∼70 nm diameter. On the other hand, 1.2 × 1021 ions m-2 s-1 flux generates a combination of randomly patched netlike nanomatrix networked structure, mostly with ∼105 nm nanostructure wall width, various-shaped pores, and self-organized nano arrays. While the observed netlike nanomatrix network structures for 8.6 × 1024 ions m-2 fluence (at a constant flux of 1.2 × 1021 ions m-2 s-1) is quite similar to those for 2.6 × 1024 ions m-2 fluence, the nanostructure wall width extends up to ∼45

PRODUCTION OF PURIFIED URANIUM

DOEpatents

Burris, L. Jr.; Knighton, J.B.; Feder, H.M.

1960-01-26

A pyrometallurgical method for processing nuclear reactor fuel elements containing uranium and fission products and for reducing uranium compound; to metallic uranium is reported. If the material proccssed is essentially metallic uranium, it is dissolved in zinc, the sulution is cooled to crystallize UZn/sub 9/ , and the UZn/sub 9/ is distilled to obtain uranium free of fission products. If the material processed is a uranium compound, the sollvent is an alloy of zinc and magnesium and the remaining steps are the same.

Recent results on the neutron irradiation of ITER candidate copper alloys irradiated in DR-3 at 250{degrees}C to 0.3 dpa

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, D.J.; Singh, B.N.; Toft, P.

1997-04-01

Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment with additional specimens re-aged and given a reactor bakeout treatment at 350{degrees}C for 100 h. CuAl-25 was also heat treated to simulate the effects of a bonding thermal cycle on the material. A number of heat treated specimens were neutron irradiated at 250{degrees}C to a dose level of {approximately}0.3 dpa in the DR-3 reactor as Riso. The main effect of the bonding thermal cycle heat treatment was a slight decrease in strength of CuCrZr and CuNiBe alloys. The strengthmore » of CuAl-25, on the other hand, remained almost unaltered. The post irradiation tests at 250{degrees}C showed a severe loss of ductility in the case of the CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens exhibited a reasonable amount of uniform elongation, with CuCrZr possessing a lower strength.« less

Surface Properties of a Nanocrystalline Fe-Ni-Nb-B Alloy After Neutron Irradiation

NASA Astrophysics Data System (ADS)

Pavùk, Milan; Sitek, Jozef; Sedlačková, Katarína

2014-09-01

The effect of neutron radiation on the surface properties of the nanocrystalline (Fe0.25Ni0.75)81Nb7B12 alloy was studied. Firstly, amorphous (Fe0.25Ni0.75)81Nb7B12 ribbon was brought by controlled annealing to the nanocrystalline state. After annealing, the samples of the nanocrystalline ribbon were irradiated in a nuclear reactor with neutron fluences of 1×1016cm-2 and 1 × 1017cm-2 . By utilizing the magnetic force microscopy (MFM), topography and a magnetic domain structure were recorded at the surface of the ribbon-shaped samples before and after irradiation with neutrons. The results indicate that in terms of surface the nanocrystalline (Fe0.25Ni0.75)81Nb7B12 alloy is radiation-resistant up to a neutron fluence of 1 × 1017cm-2 . The changes in topography observed for both irradiated samples are discussed

Effect of solute elements in Ni alloys on blistering under He + and D + ion irradiation

NASA Astrophysics Data System (ADS)

Wakai, E.; Ezawa, T.; Takenaka, T.; Imamura, J.; Tanabe, T.; Oshima, R.

2007-08-01

Effects of solute atoms on microstructural evolution and blister formation have been investigated using Ni alloys under 25 keV He + and 20 keV D + irradiation at 500 °C to a dose of about 4 × 10 21 ions/m 2. The specimens used were pure Ni, Ni-Si, Ni-Co, Ni-Cu, Ni-Mn and Ni-Pd alloys. The volume size factors of solute elements for the Ni alloys range from -5.8% to +63.6%. The formations of blisters were observed in the helium-irradiated specimens, but not in the deuteron-irradiated specimens. The areal number densities of blisters increased with volume size difference of solute atoms. The dependence of volume size on the areal number densities of blisters was very similar to that of the number densities of bubbles on solute atoms. The size of the blisters inversely decreased with increasing size of solute atoms. The formation of blisters was intimately related to the bubble growth, and the gas pressure model for the formation of blisters was supported by this study.

Castable nickel aluminide alloys for structural applications

DOEpatents

Liu, Chain T.

1992-01-01

The specification discloses nickel aluminide alloys which include as a component from about 0.5 to about 4 at. % of one or more of the elements selected from the group consisting of molybdenum or niobium to substantially improve the mechanical properties of the alloys in the cast condition.

A modified Embedded-Atom Method interatomic potential for uranium-silicide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beeler, Benjamin; Baskes, Michael; Andersson, David

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel bene ts from higher thermal conductivity and higher ssile density compared to uranium dioxide (UO 2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling e orts are underway to address this gap in knowledge. In this study, a semi-empirical modi ed Embedded-Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential is ttedmore » to the formation energy, defect energies and structural properties of U 3Si 2. The primary phase of interest (U 3Si 2) is accurately described over a wide temperature range and displays good behavior under irradiation and with free surfaces. The potential can also describe a variety of U-Si phases across the composition spectrum.« less

A modified Embedded-Atom Method interatomic potential for uranium-silicide

DOE PAGES

Beeler, Benjamin; Baskes, Michael; Andersson, David; ...

2017-08-18

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel bene ts from higher thermal conductivity and higher ssile density compared to uranium dioxide (UO 2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling e orts are underway to address this gap in knowledge. In this study, a semi-empirical modi ed Embedded-Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential is ttedmore » to the formation energy, defect energies and structural properties of U 3Si 2. The primary phase of interest (U 3Si 2) is accurately described over a wide temperature range and displays good behavior under irradiation and with free surfaces. The potential can also describe a variety of U-Si phases across the composition spectrum.« less

A modified Embedded-Atom Method interatomic potential for uranium-silicide

NASA Astrophysics Data System (ADS)

Beeler, Benjamin; Baskes, Michael; Andersson, David; Cooper, Michael W. D.; Zhang, Yongfeng

2017-11-01

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel benefits from higher thermal conductivity and higher fissile density compared to uranium dioxide (UO2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling efforts are underway to address this gap in knowledge. In this study, a semi-empirical modified Embedded-Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential is fitted to the formation energy, defect energies and structural properties of U3Si2. The primary phase of interest (U3Si2) is accurately described over a wide temperature range and displays good behavior under irradiation and with free surfaces. The potential can also describe a variety of U-Si phases across the composition spectrum.

High post-irradiation ductility thermomechanical treatment for precipitation strengthened austenitic alloys

DOEpatents

Laidler, James J.; Borisch, Ronald R.; Korenko, Michael K.

1982-01-01

A method for improving the post-irradiation ductility is described which prises a solution heat treatment following which the materials are cold worked. They are included to demonstrate the beneficial effect of this treatment on the swelling resistance and the ductility of these austenitic precipitation hardenable alloys.

Investigation of hydrogen evolution activity for the nickel, nickel-molybdenum nickel-graphite composite and nickel-reduced graphene oxide composite coatings

NASA Astrophysics Data System (ADS)

Jinlong, Lv; Tongxiang, Liang; Chen, Wang

2016-03-01

The nickel, nickel-molybdenum alloy, nickel-graphite and nickel-reduced graphene oxide composite coatings were obtained by the electrodeposition technique from a nickel sulfate bath. Nanocrystalline molybdenum, graphite and reduced graphene oxide in nickel coatings promoted hydrogen evolution reaction in 0.5 M H2SO4 solution at room temperature. However, the nickel-reduced graphene oxide composite coating exhibited the highest electrocatalytic activity for the hydrogen evolution reaction in 0.5 M H2SO4 solution at room temperature. A large number of gaps between 'cauliflower' like grains could decrease effective area for hydrogen evolution reaction in slight amorphous nickel-molybdenum alloy. The synergistic effect between nickel and reduced graphene oxide promoted hydrogen evolution, moreover, refined grain in nickel-reduced graphene oxide composite coating and large specific surface of reduced graphene oxide also facilitated hydrogen evolution reaction.

Radiation response of oxide-dispersion-strengthened alloy MA956 after self-ion irradiation

NASA Astrophysics Data System (ADS)

Chen, Tianyi; Kim, Hyosim; Gigax, Jonathan G.; Chen, Di; Wei, Chao-Chen; Garner, F. A.; Shao, Lin

2017-10-01

We studied the radiation-induced microstructural evolution of an oxide-dispersion-strengthened (ODS) ferritic alloy, MA956, to 180 dpa using 3.5 MeV Fe2+ ions. Post-irradiation examination showed that voids formed rather early and almost exclusively at the particle-matrix interfaces. Surprisingly, voids formed even in the injected interstitial zone. Comparisons with studies on other ODS alloys with smaller and largely coherent dispersoids irradiated at similar conditions revealed that the larger and not completely coherent oxide particles in MA956 serve as defect collectors which promote nucleation of voids at their interface. The interface configuration, which is related to particle type, crystal structure and size, is one of the important factors determining the defect-sink properties of particle-matrix interfaces.

The compositional dependence of irradiation creep of austenitic alloys irradiated in PFR at 420{degrees}C

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toloczko, M.B.; Garner, F.A.; Munro, B.

1997-04-01

Irradiation creep data are expensive and often difficult to obtain, especially when compared to swelling data. This requires that maximum use be made of available data sources in order to elucidate the parametric dependencies of irradiation creep for application to new alloys and to new environments such as those of proposed fusion environments. One previously untapped source of creep data is that of a joint U.S./U.K. experiment conducted in the Prototype Fast Reactor (PFR) in Dounreay, Scotland. In this experiment, five austenitic steels were irradiated in a variety of starting conditions. In particular, these steels spanned a large range (15-40%)more » of nickel contents, and contained strong variations in Mo, Ti, Al, and Nb. Some alloys were solution-strengthened and some were precipitation-strengthened. Several were cold-worked. These previously unanalyzed data show that at 420{degrees}C all austenitic steels have a creep compliance that is roughly independent of the composition of the steel at 2{+-}1 x 10{sup {minus}6}MPa{sup {minus}1} dpa{sup {minus}1}. The variation within this range may arise from the inability to completely separate the non-creep strains arising from precipitation reactions and the stress-enhancement of swelling. Each of these can be very sensitive to the composition and starting treatment of a steel.« less

PROCESS OF PREPARING URANIUM CARBIDE

DOEpatents

Miller, W.E.; Stethers, H.L.; Johnson, T.R.

1964-03-24

A process of preparing uranium monocarbide is de scribed. Uranium metal is dissolved in cadmium, zinc, cadmium-- zinc, or magnesium-- zinc alloy and a small quantity of alkali metal is added. Addition of stoichiometric amounts of carbon at 500 to 820 deg C then precipitates uranium monocarbide. (AEC)

On α‧ precipitate composition in thermally annealed and neutron-irradiated Fe- 9-18Cr alloys

NASA Astrophysics Data System (ADS)

Reese, Elaina R.; Bachhav, Mukesh; Wells, Peter; Yamamoto, Takuya; Robert Odette, G.; Marquis, Emmanuelle A.

2018-03-01

Ferritic-martensitic steels are leading candidates for many nuclear energy applications. However, formation of nanoscale α‧ precipitates during thermal aging at temperatures above 450 °C, or during neutron irradiation at lower temperatures, makes these Fe-Cr steels susceptible to embrittlement. To complement the existing literature, a series of Fe-9 to 18 Cr alloys were neutron-irradiated at temperatures between 320 and 455 °C up to doses of 20 dpa. In addition, post-irradiation annealing treatments at 500 and 600 °C were performed on a neutron-irradiated Fe-18 Cr alloy to validate the α-α‧ phase boundary. The microstructures were characterized using atom probe tomography and the results were analyzed in light of the existing literature. Under neutron irradiation and thermal annealing, the measured α‧ concentrations ranged from ∼81 to 96 at.% Cr, as influenced by temperature, precipitate size, technique artifacts, and, possibly, cascade ballistic mixing.

Characterization of faulted dislocation loops and cavities in ion irradiated alloy 800H

NASA Astrophysics Data System (ADS)

Ulmer, Christopher J.; Motta, Arthur T.

2018-01-01

Alloy 800H is a high nickel austenitic stainless steel with good high temperature mechanical properties which is considered for use in current and advanced nuclear reactor designs. The irradiation response of 800H was examined by characterizing samples that had been bulk ion irradiated at the Michigan Ion Beam Laboratory with 5 MeV Fe2+ ions to 1, 10, and 20 dpa at 440 °C. Transmission electron microscopy was used to measure the size and density of both {111} faulted dislocation loops and cavities as functions of depth from the irradiated surface. The faulted loop density increased with dose from 1 dpa up to 10 dpa where it saturated and remained approximately the same until 20 dpa. The faulted loop average diameter decreased between 1 dpa and 10 dpa and again remained approximately constant from 10 dpa to 20 dpa. Cavities were observed after irradiation doses of 10 and 20 dpa, but not after 1 dpa. The average diameter of cavities increased with dose from 10 to 20 dpa, with a corresponding small decrease in density. Cavity denuded zones were observed near the irradiated surface and near the ion implantation peak. To further understand the microstructural evolution of this alloy, FIB lift-out samples from material irradiated in bulk to 1 and 10 dpa were re-irradiated in-situ in their thin-foil geometry with 1 MeV Kr2+ ions at 440 °C at the Intermediate Voltage Electron Microscope. It was observed that the cavities formed during bulk irradiation shrank under thin-foil irradiation in-situ while dislocation loops were observed to grow and incorporate into the dislocation network. The thin-foil geometry used for in-situ irradiation is believed to cause the cavities to shrink.

Deformation-induced structural transition in body-centred cubic molybdenum

PubMed Central

Wang, S. J.; Wang, H.; Du, K.; Zhang, W.; Sui, M. L.; Mao, S. X.

2014-01-01

Molybdenum is a refractory metal that is stable in a body-centred cubic structure at all temperatures before melting. Plastic deformation via structural transitions has never been reported for pure molybdenum, while transformation coupled with plasticity is well known for many alloys and ceramics. Here we demonstrate a structural transformation accompanied by shear deformation from an original <001>-oriented body-centred cubic structure to a <110>-oriented face-centred cubic lattice, captured at crack tips during the straining of molybdenum inside a transmission electron microscope at room temperature. The face-centred cubic domains then revert into <111>-oriented body-centred cubic domains, equivalent to a lattice rotation of 54.7°, and ~15.4% tensile strain is reached. The face-centred cubic structure appears to be a well-defined metastable state, as evidenced by scanning transmission electron microscopy and nanodiffraction, the Nishiyama–Wassermann and Kurdjumov–Sachs relationships between the face-centred cubic and body-centred cubic structures and molecular dynamics simulations. Our findings reveal a deformation mechanism for elemental metals under high-stress deformation conditions. PMID:24603655

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, John P.; Miller, William E.

1989-01-01

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

Status of Post Irradiation Examination of FCAB and FCAT Irradiation Capsules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Yamamoto, Yukinori; Howard, Richard H.

A series of irradiation programs are ongoing to address the need for determining the radiation tolerance of FeCrAl alloys. These irradiation programs, deemed the FCAT and FCAB irradiation programs, use the High Flux Isotope Reactor (HFIR) to irradiate second generation wrought FeCrAl alloys and early-generation powder-metallurgy (PM) oxide dispersion-strengthened (ODS) FeCrAl alloys. Irradiations have been or are being performed at temperatures of 200°C, 330°C, and 550°C from doses of 1.8 dpa up to 16 dpa. Preliminary post-irradiation examination (PIE) on low dose (<2 dpa) irradiation capsules of tensile specimens has been performed. Analysis of co-irradiated SiC thermometry have shown reasonablemore » matching between the nominal irradiation temperatures and the target irradiation temperatures. Room temperature tensile tests have shown typical radiation-induced hardening and embrittlement at irradiations of 200°C and 330°C, but a propensity for softening when irradiated to 550°C for the wrought alloys. The PM-ODS FeCrAl specimens showed less hardening compared to the wrought alloys. Future PIE includes high temperature tensile tests on the low dose irradiation capsules as well as the determination of reference fracture toughness transition temperature, T o, in alloys irradiated to 7 dpa and higher.« less

Study of irradiation induced surface pattern and structural changes in Inconel 718 alloy

NASA Astrophysics Data System (ADS)

Wan, Hao; Si, Naichao; Zhao, Zhenjiang; Wang, Jian; Zhang, Yifei

2018-05-01

Helium ions irradiation induced surface pattern and structural changes of Inconel 718 alloy were studied with the combined utilization of atomic force microscopy (AFM), x-ray diffraction (XRD) and transmission electron microscopy (TEM). In addition, SRIM-2013 software was used to calculate the sputtering yield and detailed collision events. The result shows that, irradiation dose play an important role in altering the pattern of the surface. Enhanced irradiation aggravated the surface etching and increased the surface roughness. In ion irradiated layer, large amount of interstitials, vacancies and defect sinks were produced. Moreover, in samples with increasing dose irradiation, the dependence of interplanar spacing variation due to point defects clustering on sink density was discussed.

Carbon--silicon coating alloys for improved irradiation stability

DOEpatents

Bokros, J.C.

1973-10-01

For ceramic nuclear fuel particles, a fission product-retaining carbon-- silicon alloy coating is described that exhibits low shrinkage after exposure to fast neutron fluences of 1.4 to 4.8 x 10/sup 21/ n/cm/sup 2/ (E = 0.18 MeV) at irradiation temperatures from 950 to 1250 deg C. Isotropic pyrolytic carbon containing from 18 to 34 wt% silicon is co-deposited from a gaseous mixiure of propane, helium, and silane at a temperature of 1350 to 1450 deg C. (Official Gazette)

Advances in Low Carbon, High Strength Ferrous Alloys

DTIC Science & Technology

1993-04-01

35 TABLES 1. Specified chemical compositions and mechanical properties for GMAW/SAW/ GTAW wire electrodes, MIL-XXXS type, for welding...minimum service temperature of +300 F. The chromium and molybdenum additions improved hardenability and promoted the formation of mar- tensite in thick...alloying ele- ments ( chromium , nickel and molybdenum) are required, especially for thick sections. Production of high strength steel plate for military

Vapor deposition of molybdenum oxide using bis(ethylbenzene) molybdenum and water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Drake, Tasha L.; Stair, Peter C., E-mail: pstair@u.northwestern.edu

2016-09-15

Three molybdenum precursors—bis(acetylacetonate) dioxomolybdenum, molybdenum isopropoxide, and bis(ethylbenzene) molybdenum—were tested for molybdenum oxide vapor deposition. Quartz crystal microbalance studies were performed to monitor growth. Molybdenum isopropoxide and bis(ethylbenzene) molybdenum achieved linear growth rates 0.01 and 0.08 Å/cycle, respectively, using atomic layer deposition techniques. Negligible MoO{sub x} growth was observed on alumina powder using molybdenum isopropoxide, as determined by inductively coupled plasma optical emission spectroscopy. Bis(ethylbenzene) molybdenum achieved loadings of 0.5, 1.1, and 1.9 Mo/nm{sup 2} on alumina powder after one, two, and five cycles, respectively, using atomic layer deposition techniques. The growth window for bis(ethylbenzene) molybdenum is 135–150 °C. An alternative pulsingmore » strategy was also developed for bis(ethylbenzene) molybdenum that results in higher growth rates in less time compared to atomic layer deposition techniques. The outlined process serves as a methodology for depositing molybdenum oxide for catalytic applications. All as-deposited materials undergo further calcination prior to characterization and testing.« less

Forming a structure of the CoNiFe alloys by X-ray irradiation

NASA Astrophysics Data System (ADS)

Valko, Natalia; Kasperovich, Andrey; Koltunowicz, Tomasz N.

The experimental data of electrodeposition kinetics researches and structure formation of ternary CoNiFe alloys deposited onto low-carbon steel 08kp in the presence of X-rays are presented. Relations of deposit rate, current efficiencies, element and phase compositions of CoNiFe coatings formed from sulfate baths with respect to cathode current densities (0.5-3A/dm2), electrolyte composition and irradiation were obtained. It is shown that, the CoNiFe coatings deposited by the electrochemical method involving exposure of the X-rays are characterized by more perfect morphology surfaces with less developed surface geometry than reference coatings. The effect of the X-ray irradiation on the electrodeposition of CoNiFe coatings promotes formatting of alloys with increased electropositive component and modified phase composition.

Effect of 0.25 and 2.0 MeV He-Ion Irradiation on Short-Range Ordering in Model (EFDA) Fe-Cr Alloys

NASA Astrophysics Data System (ADS)

Dubiel, Stanisław M.; Żukrowski, Jan; Serruys, Yves

2018-05-01

The effects of He+ irradiation on a distribution of Cr atoms in Fe100-x Cr x (x = 5.8, 10.75, 15.15) alloys were studied by 57Fe Conversion Electron Mössbauer Spectroscopy (CEMS). The alloys were irradiated with doses up to 12 × 1016 ions/cm2 with 0.25 and 2.0 MeV He+ ions. The distribution of Cr atoms within the first two coordination shells around Fe atoms was expressed with short-range order parameters α 1 (first-neighbor shell, 1NN), α 2 (second-neighbor shell, 2NN), and α 12 (1NN + 2NN). In non-irradiated alloys, α 1 >0 and α 2 <0 was revealed for all three samples. The value of α 12 ≈0, i.e., the distribution of Cr atoms averaged over 1NN and 2NN, was random. The effect of the irradiation of the Fe94.2Cr5.8 alloy was similar for the two energies of He+, viz., increase of number of Cr atoms in 1NN and decrease in 2NN. Consequently, the degree of ordering increased. For the other two samples, the effect of the irradiation depends on the composition, and is stronger for the less energetic ions where, for Fe89.25Cr10.75 alloy, the disordering disappeared and some traces of Cr clustering appeared. In Fe84.85Cr15.15 alloy, the clustering was clear. In the samples irradiated with 2. 0 MeV He+ ions, the ordering also survived in the samples with x = 10.75 and 15.15, yet its degree became smaller than in the Fe94.2Cr5.8 alloy.

Processing of LEU targets for {sup 99}Mo production--testing and modification of the Cintichem process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wu, D.; Landsberger, S.; Buchholz, B.

1995-09-01

Recent experimental results on testing and modification of the Cintichem process to allow substitution of low enriched uranium (LEU) for high enriched uranium (HEU) targets are presented in this report. The main focus is on {sup 99}Mo recovery and purification by its precipitation with {alpha}-benzoin oxime. Parameters that were studied include concentrations of nitric and sulfuric acids, partial neutralization of the acids, molybdenum and uranium concentrations, and the ratio of {alpha}-benzoin oxime to molybdenum. Decontamination factors for uranium, neptunium, and various fission products were measured. Experiments with tracer levels of irradiated LEU were conducted for testing the {sup 99}Mo recoverymore » and purification during each step of the Cintichem process. Improving the process with additional processing steps was also attempted. The results indicate that the conversion of molybdenum chemical processing from HEU to LEU targets is possible.« less

Establishment of Wear Resistant HVOF Coatings for 50CrMo4 Chromium Molybdenum Alloy Steel as an Alternative for Hard Chrome Plating

NASA Astrophysics Data System (ADS)

Karuppasamy, S.; Sivan, V.; Natarajan, S.; Kumaresh Babu, S. P.; Duraiselvam, M.; Dhanuskodi, R.

2018-05-01

High cost imported components of seamless steel tube manufacturing plants wear frequently and need replacement to ensure the quality of the product. Hard chrome plating, which is time consuming and hazardous, is conventionally used to restore the original dimension of the worn-out surface of the machine components. High Velocity Oxy-Fuel (HVOF) thermal spray coatings with NiCrBSi super alloy powder and Cr3C2 NiCr75/25 alloy powder applied on a 50CrMo4 (DIN-1.7228) chromium molybdenum alloy steel, the material of the wear prone machine component, were evaluated for use as an alternative for hard chrome plating in this present work. The coating characteristics are evaluated using abrasive wear test, sliding wear test and microscopic analysis, hardness test, etc. The study results revealed that the HVOF based NiCrBSi and Cr3C2NiCr75/25 coatings have hardness in the range of 800-900 HV0.3, sliding wear rate in the range of 50-60 µm and surface finish around 5 microns. Cr3C2 NiCr75/25 coating is observed to be a better option out of the two coatings evaluated for the selected application.

Process for removing carbon from uranium

DOEpatents

Powell, George L.; Holcombe, Jr., Cressie E.

1976-01-01

Carbon contamination is removed from uranium and uranium alloys by heating in inert atmosphere to 700.degree.-1900.degree.C in effective contact with yttrium to cause carbon in the uranium to react with the yttrium. The yttrium is either in direct contact with the contaminated uranium or in indirect contact by means of an intermediate transport medium.

Undercooled and rapidly quenched Ni-Mo alloys

NASA Technical Reports Server (NTRS)

Tewari, S. N.; Glasgow, T. K.

1986-01-01

Hypoeutectic, eutectic, and hypereutectic nickel-molybdenum alloys were rapidly solidified by both bulk undercooling and melt spinning techniques. Alloys were undercooled in both electromagnetic levitation and differential thermal analysis equipment. The rate of recalescence depended upon the degree of initial undercooling and the nature (faceted or nonfaceted) of the primary nucleating phase. Alloy melts were observed to undercool more in the presence of primary Beta (NiMo intermetallic) phase than in gamma (fcc solid solution) phase. Melt spinning resulted in an extension of molybdenum solid solubility in gamma nickel, from 28 to 37.5 at % Mo. Although the microstructures observed by undercooling and melt spinning were similar the microsegregation pattern across the gamma dendries was different. The range of microstructures evolved was analyzed in terms of the nature of the primary phase to nucleate, its subsequent dendritic growth, coarsening and fragmentation, and final solidification of interfenderitic liquid.

Monte Carlo Criticality Analysis of Simple Geometrics COntaining Tungsten Rhenium Alloys Engrained with Uranium Dioxide and Uranium Mononitride

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jonathan A. Webb; Indrajit Charit

2011-08-01

The critical mass and dimensions of simple geometries containing highly enriched uraniumdioxide (UO2) and uraniummononitride (UN) encapsulated in tungsten-rhenium alloys are determined using MCNP5 criticality calculations. Spheres as well as cylinders with length to radius ratios of 1.82 are computationally built to consist of 60 vol.% fuel and 40 vol.% metal matrix. Within the geometries the uranium is enriched to 93 wt.% uranium-235 and the rhenium content within the metal alloy was modeled over a range of 0 to 30 at.%. The spheres containing UO2 were determined to have a critical radius of 18.29 cm to 19.11 cm and amore » critical mass ranging from 366 kg to 424 kg. The cylinders containing UO2 were found to have a critical radius ranging from 17.07 cm to 17.844 cm with a corresponding critical mass of 406 kg to 471 kg. Spheres engrained with UN were determined to have a critical radius ranging from 14.82 cm to 15.19 cm and a critical mass between 222 kg and 242 kg. Cylinders which were engrained with UN were determined to have a critical radius ranging from 13.811 cm to 14.155 cm with a corresponding critical mass of 245 kg to 267 kg. The critical geometries were also computationally submerged in a neutronaically infinite medium of fresh water to determine the effects of rhenium addition on criticality accidents due to water submersion. The monte carlo analysis demonstrated that rhenium addition of up to 30 at.% can reduce the excess reactivity due to water submersion by up to $5.07 for UO2 fueled cylinders, $3.87 for UO2 fueled spheres and approximately $3.00 for UN fueled spheres and cylinders.« less

Microstructural stability and mechanical behavior of FeNiMnCr high entropy alloy under ion irradiation

DOE PAGES

Leonard, Keith J.; Bei, Hongbin; Zinkle, Steven J.; ...

2016-05-13

In recent years, high entropy alloys (HEAs) have attracted significant attention due to their excellent mechanical properties and good corrosion resistance, making them potential candidates for high temperature fission and fusion structural applications. However there is very little known about their radiation resistance, particularly at elevated temperatures relevant for energy applications. In the present study, a single phase (face centered cubic) concentrated solid solution alloy of composition 27%Fe-28%Ni-27%Mn-18%Cr was irradiated with 3 or 5.8 MeV Ni ions at temperatures ranging from room temperature to 700 °C and midrange doses from 0.03 to 10 displacements per atom (dpa). Transmission electron microscopymore » (TEM), scanning transmission electron microscopy with energy dispersive x-ray spectrometry (STEM/EDS) and X-ray diffraction (XRD) were used to characterize the radiation defects and microstructural changes. Irradiation at higher temperatures showed evidence of relatively sluggish solute diffusion with limited solute depletion or enrichment at grain boundaries. The main microstructural feature at all temperatures was high-density small dislocation loops. Voids were not observed at any irradiation condition. Nano-indentation tests on specimens irradiated at room temperature showed a rapid increase in hardness ~35% and ~80% higher than the unirradiated value at 0.03 and 0.3 dpa midrange doses, respectively. The irradiation-induced hardening was less pronounced for 500 °C irradiations (<20% increase after 3 dpa). Overall, the examined HEA material exhibits superior radiation resistance compared to conventional single phase Fe-Cr-Ni austenitic alloys such as stainless steels. Furthermore, the present study provides insight on the fundamental irradiation behavior of a single phase HEA material over a broad range of irradiation temperatures.« less

METHOD AND FLUX COMPOSITION FOR TREATING URANIUM

DOEpatents

Foote, F.

1958-08-23

ABS>A flux composition is described fer use with molten uranium or uranium alloys. The flux consists of about 46 weight per cent calcium fiuoride, 46 weight per cent magnesium fluoride and about 8 weight per cent of uranium tetrafiuoride.

Quantitative experimental determination of the solid solution hardening potential of rhenium, tungsten and molybdenum in single-crystal nickel-based superalloys

DOE PAGES

Fleischmann, Ernst; Miller, Michael K.; Affeldt, Ernst; ...

2015-01-31

Here, the solid-solution hardening potential of the refractory elements rhenium, tungsten and molybdenum in the matrix of single-crystal nickel-based superalloys was experimentally quantified. Single-phase alloys with the composition of the nickel solid-solution matrix of superalloys were cast as single crystals, and tested in creep at 980 °C and 30–75 MPa. The use of single-phase single-crystalline material ensures very clean data because no grain boundary or particle strengthening effects interfere with the solid-solution hardening. This makes it possible to quantify the amount of rhenium, tungsten and molybdenum necessary to reduce the creep rate by a factor of 10. Rhenium is moremore » than two times more effective for matrix strengthening than either tungsten or molybdenum. The existence of rhenium clusters as a possible reason for the strong strengthening effect is excluded as a result of atom probe tomography measurements. If the partitioning coefficient of rhenium, tungsten and molybdenum between the γ matrix and the γ' precipitates is taken into account, the effectiveness of the alloying elements in two-phase superalloys can be calculated and the rhenium effect can be explained.« less

Surface modifications of hydrogen storage alloy by heavy ion beams with keV to MeV irradiation energies

NASA Astrophysics Data System (ADS)

Abe, Hiroshi; Tokuhira, Shinnosuke; Uchida, Hirohisa; Ohshima, Takeshi

2015-12-01

This study deals with the effect of surface modifications induced from keV to MeV heavy ion beams on the initial reaction rate of a hydrogen storage alloy (AB5) in electrochemical process. The rare earth based alloys like this sample alloy are widely used as a negative electrode of Ni-MH (Nickel-Metal Hydride) battery. We aimed to improve the initial reaction rate of hydrogen absorption by effective induction of defects such as vacancies, dislocations, micro-cracks or by addition of atoms into the surface region of the metal alloys. Since defective layer near the surface can easily be oxidized, the conductive oxide layer is formed on the sample surface by O+ beams irradiation, and the conductive oxide layer might cause the improvement of initial reaction rate of hydriding. This paper demonstrates an effective surface treatment of heavy ion irradiation, which induces catalytic activities of rare earth oxides in the alloy surface.

Determination of small amounts of molybdenum in tungsten and molybdenum ores

USGS Publications Warehouse

Grimaldi, F.S.; Wells, R.C.

1943-01-01

A rapid method has been developed for the determination of small amounts of molybdenum in tungsten and molybdenum ores. After removing iron and other major constituents the molybdenum thiocyanate color is developed in water-acetone solutions, using ammonium citrate to eliminate the interference of tungsten. Comparison is made by titrating a blank with a standard molybdenum solution. Aliquots are adjusted to deal with amounts of molybdenum ranging from 0.01 to 1.30 mg.

PROCESS OF DISSOLVING ZIRCONIUM ALLOYS

DOEpatents

Shor, R.S.; Vogler, S.

1958-01-21

A process is described for dissolving binary zirconium-uranium alloys where the uranium content is about 2%. In prior dissolution procedures for these alloys, an oxidizing agent was added to prevent the precipitation of uranium tetrafluoride. In the present method complete dissolution is accomplished without the use of the oxidizing agent by using only the stoichiometric amount or slight excess of HF required by the zirconium. The concentration of the acid may range from 2M to 10M and the dissolution is advatageously carried out at a temperature of 80 deg C.

Precipitate resolution in an electron irradiated ni-si alloy

NASA Astrophysics Data System (ADS)

Watanabe, H.; Muroga, T.; Yoshida, N.; Kitajima, K.

1988-09-01

Precipitate resolution processes in a Ni-12.6 at% Si alloy under electron irradiation have been observed by means of HVEM. Above 400°C, growth and resolution of Ni 3Si precipitates were observed simultaneously. The detail stereoscopic observation showed that the precipitates close to free surfaces grew, while those in the middle of a specimen dissolved. The critical dose when the precipitates start to shrink increases with increasing the depth. This depth dependence of the precipitate behavior under irradiation has a close relation with the formation of surface precipitates and the growth of solute depleted zone beneath them. The temperature and dose dependence of the resolution rate showed that the precipitates in the solute depleted zone dissolved by the interface controlled process of radiation-enhanced diffusion.

Influence of the Additives and The pH On the Cobalt-Molybdenum (Co-Mo) Alloy Electrodeposited On n-TypeSilicon

NASA Astrophysics Data System (ADS)

Fekih, Z.; Ghellai, N.; Fortas, G.; Chiboub, N.; Sam, S.; Chabanne-sari, N. E.; Gabouze, N.

In this work, thin films of metal alloys (Co-Mo) have been electrodeposited onto silicon (Si) surface. The effects of two different additives (H3BO3 and Na2CO3) and the pH of the solution on the electrochemically deposited films (morphology, stochiometry…) have been investigated. The properties of the deposits were characterized by using X-Rays Diffraction (XRD), Scanning Electron Microscopy (SEM) and Energy Dispersive X-ray Spectroscopy (EDS). The results show that the morphology and the film composition depend on both the pH of the solution and the additives. The presence of boric acid favors the Mo deposition. Crack-free homogeneous deposits with a low percentage of molybdenum can be easily obtained from high pH bath. The deposits were shown to exhibits a good crystalline structure.

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

DOE PAGES

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth; ...

2018-03-30

Here, we demonstrate here that locally stabilized structure and compositional segregation at grain boundaries in a complex multicomponent alloy can be modified using high temperature homogenization treatment to influence the kinetics of phase transformations initiating from grain boundaries during subsequent low temperature annealing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography of a model multicomponent metallic alloy —uranium-10 wt% molybdenum (U-10Mo) a nuclear fuel, that is highly relevant to worldwide nuclear non-proliferation efforts, we demonstrate the ability to change the structure and compositional segregation at grain boundary, which then controls the subsequent discontinuous precipitation kinetics during sub-eutectoid annealing.more » A change in grain boundary from one characterized by segregation of Mo and impurities at grain boundary to a phase boundary with a distinct U 2MoSi 2C wetting phase precipitates introducing Ni and Al rich interphase complexions caused a pronounced reduction in area fraction of subsequent discontinuous precipitation. The broader implication of this work is in highlighting the role of grain boundary structure and composition in metallic alloys on dictating the fate of grain boundary initiated phase transformations like discontinuous precipitation or cellular transformation. This work highlights a new pathway to tune the grain boundary structure and composition to tailor the final microstructure of multicomponent metallic alloys.« less

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth

Here, we demonstrate here that locally stabilized structure and compositional segregation at grain boundaries in a complex multicomponent alloy can be modified using high temperature homogenization treatment to influence the kinetics of phase transformations initiating from grain boundaries during subsequent low temperature annealing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography of a model multicomponent metallic alloy —uranium-10 wt% molybdenum (U-10Mo) a nuclear fuel, that is highly relevant to worldwide nuclear non-proliferation efforts, we demonstrate the ability to change the structure and compositional segregation at grain boundary, which then controls the subsequent discontinuous precipitation kinetics during sub-eutectoid annealing.more » A change in grain boundary from one characterized by segregation of Mo and impurities at grain boundary to a phase boundary with a distinct U 2MoSi 2C wetting phase precipitates introducing Ni and Al rich interphase complexions caused a pronounced reduction in area fraction of subsequent discontinuous precipitation. The broader implication of this work is in highlighting the role of grain boundary structure and composition in metallic alloys on dictating the fate of grain boundary initiated phase transformations like discontinuous precipitation or cellular transformation. This work highlights a new pathway to tune the grain boundary structure and composition to tailor the final microstructure of multicomponent metallic alloys.« less

Molybdenum in the environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jawell, W.M.; Page, A.L.; Elseewi, A.A.

1980-01-01

While molybdenum is an essential element for both plants and animals, it becomes toxic above certain critical levels. Reviewed are the natural supply of molybdenum in the environment. The molybdenum cycle, the importance of molybdenum in industry and agriculture, and potential hazards that may occur when excessive levels of molybdenum occur in the environment. Although the potential of molybdenum toxicity to humans and non-ruminant animals appears to be low, the enrichment of the environment with molybdenum from modern mining, agricultural, and industrial activities has potentially hazardous implications for ruminant animal health.

Molybdenum in the environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jarrell, W.M.; Page, A.L.; Elseewi, A.A.

1980-01-01

While molybdenum is an essential element for both plants and animals, it becomes toxic above certain critical levels. Reviewed are the natural supply of molybdenum in the environment, the molybdenum cycle, the importance of molybdenum in industry and agriculture, and potential hazards that may occur when excessive levels of molybdenum occur in the environment. Although the potential of molybdenum toxicity to humans and non-ruminant animals appears to be low, the enrichment of the environment with molybdenum from modern mining, agricultural, and industrial activities has potentially hazardous implications for ruminant animal health. (3 graphs, numerous references, 16 tables)

Phase stability and microstructures of high entropy alloys ion irradiated to high doses

NASA Astrophysics Data System (ADS)

Xia, Songqin; Gao, Michael C.; Yang, Tengfei; Liaw, Peter K.; Zhang, Yong

2016-11-01

The microstructures of AlxCoCrFeNi (x = 0.1, 0.75 and 1.5 in molar ratio) high entropy alloys (HEAs) irradiated at room temperature with 3 MeV Au ions at the highest fluence of 105, 91, and 81 displacement per atom, respectively, were studied. Transmission electron microscopy (TEM) and high-resolution TEM (HRTEM) analyses show that the initial microstructures and phase composition of all three alloys are retained after ion irradiation and no phase decomposition is observed. Furthermore, it is demonstrated that the disordered face-centered cubic (FCC) and disordered body-centered cubic (BCC) phases show much less defect cluster formation and structural damage than the NiAl-type ordered B2 phase. This effect is explained by higher entropy of mixing, higher defect formation/migration energies, substantially lower thermal conductivity, and higher atomic level stress in the disordered phases.

Preliminary Analysis of the General Performance and Mechanical Behavior of Irradiated FeCrAl Base Alloys and Weldments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gussev, Maxim N.; Field, Kevin G.; Briggs, Samuel A.

The iron-based, iron-chromium-aluminum (FeCrAl) alloys are promising, robust materials for deployment in current and future nuclear power plants. This class of alloys demonstrates excellent performance in a range of environments and conditions, including high-temperature steam (>1000°C). Furthermore, these alloys have the potential to have prolonged survival under loss-of-coolant accident (LOCA) conditions compared to the more traditional cladding materials that are either Zr-based alloys or austenitic steels. However, one of the issues associated with FeCrAl alloys is cracking during welding. The present project investigates the possibility of mitigating welding-induced cracking via alloying and precise structure control of the weldments; in themore » frame work of the project, several advanced alloys were developed and are being investigated prior to and after neutron irradiation to provide insight into the radiation tolerance and mechanical performance of the weldments. The present report provides preliminary results on the post-irradiation characterization and mechanical tests performed during United States Fiscal Year (FY) 2016. Chapter 1 provides a general introduction, and Chapter 2 describes the alloy compositions, welding procedure, specimen geometry and manufacturing parameters. Also, a brief discussion of the irradiation at the High Flux Isotope Reactor (HFIR) is provided. Chapter 3 is devoted to the analysis of mechanical tests performed at the hot cell facility; tensile curves and mechanical properties are discussed in detail focusing on the irradiation temperature. Limited fractography results are also presented and analyzed. The discussion highlights the limitations of the testing within a hot cell. Chapter 4 underlines the advantages of in-situ testing and discusses the preliminary results obtained with newly developed miniature specimens. Specimens were moved to the Low Activation Materials Development and Analysis (LAMDA) laboratory and

The effect of the initial microstructure in terms of sink strength on the ion-irradiation-induced hardening of ODS alloys studied by nanoindentation

NASA Astrophysics Data System (ADS)

Duan, Binghuang; Heintze, Cornelia; Bergner, Frank; Ulbricht, Andreas; Akhmadaliev, Shavkat; Oñorbe, Elvira; de Carlan, Yann; Wang, Tieshan

2017-11-01

Oxide dispersion strengthened (ODS) Fe-Cr alloys are promising candidates for structural components in nuclear energy production. The small grain size, high dislocation density and the presence of particle matrix interfaces may contribute to the improved irradiation resistance of this class of alloys by providing sinks and/or traps for irradiation-induced point defects. The extent to which these effects impede hardening is still a matter of debate. To address this problem, a set of alloys of different grain size, dislocation density and oxide particle distribution were selected. In this study, three-step Fe-ion irradiation at both 300 °C and 500 °C up to 10 dpa was used to introduce damage in five different materials including three 9Cr-ODS alloys, one 14Cr-ODS alloy and one 14Cr-non-ODS alloy. Electron backscatter diffraction (EBSD), transmission electron microscopy (TEM), small angle neutron scattering (SANS), and nanoindentation testing were applied, the latter before and after irradiation. Significant hardening occurred for all materials and temperatures, but it is distinctly lower in the 14Cr alloys and also tends to be lower at the higher temperature. The possible contribution of Cr-rich α‧-phase particles is addressed. The impact of grain size, dislocation density and particle distribution is demonstrated in terms of an empirical trend between total sink strength and hardening.

Method of preparation of uranium nitride

DOEpatents

Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

2013-07-09

Method for producing terminal uranium nitride complexes comprising providing a suitable starting material comprising uranium; oxidizing the starting material with a suitable oxidant to produce one or more uranium(IV)-azide complexes; and, sufficiently irradiating the uranium(IV)-azide complexes to produce the terminal uranium nitride complexes.

Ultrasonic irradiation and its application for improving the corrosion resistance of phosphate coatings on aluminum alloys.

PubMed

Sheng, Minqi; Wang, Chao; Zhong, Qingdong; Wei, Yinyin; Wang, Yi

2010-01-01

In this paper, ultrasonic irradiation was utilized for improving the corrosion resistance of phosphate coatings on aluminum alloys. The chemical composition and morphology of the coatings were analyzed by X-ray diffraction analysis (XRD) and scanning electron microscopy (SEM). The effect of ultrasonic irradiation on the corrosion resistance of phosphate coatings was investigated by polarization curves and electrochemical impedance spectroscopy (EIS). Various effects of the addition of Nd(2)O(3) in phosphating bath on the performance of the coatings were also investigated. Results show that the composition of phosphate coating were Zn(3)(PO(4))(2).4H(2)O(hopeite) and Zn crystals. The phosphate coatings became denser with fewer microscopic holes by utilizing ultrasonic irradiation treatment. The addition of Nd(2)O(3) reduced the crystallinity of the coatings, with the additional result that the crystallites were increasingly nubby and spherical. The corrosion resistance of the coatings was also significantly improved by ultrasonic irradiation treatment; both the anodic and cathodic processes of corrosion taking place on the aluminum alloy substrate were suppressed consequently. In addition, the electrochemical impedance of the coatings was also increased by utilizing ultrasonic irradiation treatment compared with traditional treatment.

Production of medical isotopes from a thorium target irradiated by light charged particles up to 70 MeV

NASA Astrophysics Data System (ADS)

Duchemin, C.; Guertin, A.; Haddad, F.; Michel, N.; Métivier, V.

2015-02-01

The irradiation of a thorium target by light charged particles (protons and deuterons) leads to the production of several isotopes of medical interest. Direct nuclear reaction allows the production of Protactinium-230 which decays to Uranium-230 the mother nucleus of Thorium-226, a promising isotope for alpha radionuclide therapy. The fission of Thorium-232 produces fragments of interest like Molybdenum-99, Iodine-131 and Cadmium-115g. We focus our study on the production of these isotopes, performing new cross section measurements and calculating production yields. Our new sets of data are compared with the literature and the last version of the TALYS code.

Production of medical isotopes from a thorium target irradiated by light charged particles up to 70 MeV.

PubMed

Duchemin, C; Guertin, A; Haddad, F; Michel, N; Métivier, V

2015-02-07

The irradiation of a thorium target by light charged particles (protons and deuterons) leads to the production of several isotopes of medical interest. Direct nuclear reaction allows the production of Protactinium-230 which decays to Uranium-230 the mother nucleus of Thorium-226, a promising isotope for alpha radionuclide therapy. The fission of Thorium-232 produces fragments of interest like Molybdenum-99, Iodine-131 and Cadmium-115g. We focus our study on the production of these isotopes, performing new cross section measurements and calculating production yields. Our new sets of data are compared with the literature and the last version of the TALYS code.

New alloys to conserve critical elements

NASA Technical Reports Server (NTRS)

Stephens, J. R.

1978-01-01

Based on availability of domestic reserves, chromium is one of the most critical elements within the U.S. metal industry. New alloys having reduced chromium contents which offer potential as substitutes for higher chromium containing alloys currently in use are being investigated. This paper focuses primarily on modified Type 304 stainless steels having one-third less chromium, but maintaining comparable oxidation and corrosion properties to that of type 304 stainless steel, the largest single use of chromium. Substitutes for chromium in these modified Type 304 stainless steel alloys include silicon and aluminum plus molybdenum.

Nickel aluminide alloy suitable for structural applications

DOEpatents

Liu, Chain T.

1998-01-01

Alloys for use in structural applications based upon NiAl to which are added selected elements to enhance room temperature ductility and high temperature strength. Specifically, small additions of molybdenum produce a beneficial alloy, while further additions of boron, carbon, iron, niobium, tantalum, zirconium and hafnium further improve performance of alloys at both room temperature and high temperatures. A preferred alloy system composition is Ni--(49.1.+-.0.8%)Al--(1.0.+-.0.8%)Mo--(0.7.+-.0.5%)Nb/Ta/Zr/Hf--(nearly zero to 0.03%)B/C, where the % is at. % in each of the concentrations. All alloys demonstrated good oxidation resistance at the elevated temperatures. The alloys can be fabricated into components using conventional techniques.

Investigation of americium-241 metal alloys for target applications. [Alloys with cerium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Conner, W.V.

1980-01-01

Several americium-241 metal alloys have been investigated for possible use in the Lawrence Livermore National Laboratory Radiochemical Diagnostic Tracer Program. Alloys investigated have included uranium-americium, aluminum-americium, and cerium-americium. Uranium-americium alloys with the desired properties proved to be difficult to prepare, and work with this alloy was discontinued. Aluminum-americium alloys were much easier to prepare, but the alloy consisted of an aluminum-americium intermetallic compound (AmAl/sub 4/) in an aluminum matrix. This alloy could be cast and formed into shapes, but the low density of aluminum, and other problems; made the alloy unsuitable for the intended application. Americium metal was found tomore » have a high solid solubility in cerium and alloys prepared from these two elements exhibited all of the properties desired for the tracer program application. Cerium-americium alloys containing up to 34 wt % americium have been prepared using both comelting and coreduction techniques. The latter technique involves coreduction of Ce F/sub 4/ and AmF/sub 4/ with calcium metal in a sealed reduction vessel. Casting techniques have been developed for preparing up to eight 0.87 inch (2.2 cm) diameter disks in a single casting, and cerium-americium metal alloy disks containing from 10 to 25 wt % americium-241 have been prepared using these techniques.« less

Defect structures induced by high-energy displacement cascades in γ uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miao, Yinbin; Beeler, Benjamin; Deo, Chaitanya

Displacement cascade simulations were conducted for the c uranium system based on molecular dynamics. A recently developed modified embedded atom method (MEAM) potential was employed to replicate the atomic interactions while an embedded atom method (EAM) potential was adopted to help characterize the defect structures induced by the displacement cascades. The atomic displacement process was studied by providing primary knock-on atoms (PKAs) with kinetic energies from 1 keV to 50 keV. The influence of the PKA incident direction was examined. The defect structures were analyzed after the systems were fully relaxed. The states of the self-interstitial atoms (SIAs) were categorizedmore » into various types of dumbbells, the crowdion, and the octahedral interstitial. The voids were determined to have a polyhedral shape with {110} facets. The size distribution of the voids was also obtained. The results of this study not only expand the knowledge of the microstructural evolution in irradiated c uranium, but also provide valuable references for the radiation-induced defects in uranium alloy fuels.« less

Low Temperature Diffusion Transformations in Fe-Ni-Ti Alloys During Deformation and Irradiation

NASA Astrophysics Data System (ADS)

Sagaradze, Victor; Shabashov, Valery; Kataeva, Natalya; Kozlov, Kirill; Arbuzov, Vadim; Danilov, Sergey; Ustyugov, Yury

2018-03-01

The deformation-induced dissolution of Ni3Ti intermetallics in the matrix of austenitic alloys of Fe-36Ni-3Ti type was revealed in the course of their cascade-forming neutron irradiation and cold deformation at low temperatures via employment of Mössbauer method. The anomalous deformation-related dissolution of the intermetallics has been explained by the migration of deformation-induced interstitial atoms from the particles into a matrix in the stress field of moving dislocations. When rising the deformation temperature, this process is substituted for by the intermetallics precipitation accelerated by point defects. A calculation of diffusion processes has shown the possibility of the realization of the low-temperature diffusion of interstitial atoms in configurations of the crowdions and dumbbell pairs at 77-173 K. The existence of interstitial atoms in the Fe-36Ni alloy irradiated by electrons or deformed at 77 K was substantiated in the experiments of the electrical resistivity measurements.

Molybdenum disilicide composites

DOEpatents

Rodriguez, Robert P.; Petrovic, John J.

2001-01-01

Molybdenum disilicide/.beta.'-Si.sub.6-z Al.sub.z O.sub.z N.sub.8-z, wherein z=a number from greater than 0 to about 5, composites are made by use of in situ reactions among .alpha.-silicon nitride, molybdenum disilicide, and aluminum. Molybdenum disilicide within a molybdenum disilicide/.beta.'-Si.sub.6-z Al.sub.z O.sub.z N.sub.8-z eutectoid matrix is the resulting microstructure when the invention method is employed.

High-Resolution Characterization of UMo Alloy Microstructure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Devaraj, Arun; Kovarik, Libor; Joshi, Vineet V.

2016-11-30

This report highlights the capabilities and procedure for high-resolution characterization of UMo fuels in PNNL. Uranium-molybdenum (UMo) fuel processing steps, from casting to forming final fuel, directly affect the microstructure of the fuel, which in turn dictates the in-reactor performance of the fuel under irradiation. In order to understand the influence of processing on UMo microstructure, microstructure characterization techniques are necessary. Higher-resolution characterization techniques like transmission electron microscopy (TEM) and atom probe tomography (APT) are needed to interrogate the details of the microstructure. The findings from TEM and APT are also directly beneficial for developing predictive multiscale modeling tools thatmore » can predict the microstructure as a function of process parameters. This report provides background on focused-ion-beam–based TEM and APT sample preparation, TEM and APT analysis procedures, and the unique information achievable through such advanced characterization capabilities for UMo fuels, from a fuel fabrication capability viewpoint.« less

Modeling copper precipitation hardening and embrittlement in a dilute Fe-0.3at.%Cu alloy under neutron irradiation

NASA Astrophysics Data System (ADS)

Bai, Xian-Ming; Ke, Huibin; Zhang, Yongfeng; Spencer, Benjamin W.

2017-11-01

Neutron irradiation in light water reactors can induce precipitation of nanometer sized Cu clusters in reactor pressure vessel steels. The Cu precipitates impede dislocation gliding, leading to an increase in yield strength (hardening) and an upward shift of ductile-to-brittle transition temperature (embrittlement). In this work, cluster dynamics modeling is used to model the entire Cu precipitation process (nucleation, growth, and coarsening) in a Fe-0.3at.%Cu alloy under neutron irradiation at 300°C based on the homogenous nucleation mechanism. The evolution of the Cu cluster number density and mean radius predicted by the modeling agrees well with experimental data reported in literature for the same alloy under the same irradiation conditions. The predicted precipitation kinetics is used as input for a dispersed barrier hardening model to correlate the microstructural evolution with the radiation hardening and embrittlement in this alloy. The predicted radiation hardening agrees well with the mechanical test results in the literature. Limitations of the model and areas for future improvement are also discussed in this work.

Effects of irradiation to 4 dpa at 390 C on the fracture toughness of vanadium alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gruber, E.E.; Galvin, T.M.; Chopra, O.K.

Fracture toughness J-R curve tests were conducted at room temperature on disk-shaped compact-tension DC(T) specimens of three vanadium alloys having a nominal composition of V-4Cr-4Ti. The alloys in the nonirradiated condition showed high fracture toughness; J{sub IC} could not be determined but is expected to be above 600 kJ/m{sup 2}. The alloys showed very poor fracture toughness after irradiation to 4 dpa at 390 C, e.g., J{sub IC} values of {approx}10 kJ/m{sup 2} or lower.

Development of Nitride Coating Using Atomic Layer Deposition for Low-Enriched Uranium Fuel Powder

NASA Astrophysics Data System (ADS)

Bhattacharya, Sumit

High-performance research reactors require fuel that operates at high specific power and can withstand high fission density, but at relatively low temperatures. The design of the research reactor fuels is done for efficient heat emission, and consists of assemblies of thin-plates cladding made from aluminum alloy. The low-enriched fuels (LEU) were developed for replacing high-enriched fuels (HEU) for these reactors necessitates a significantly increased uranium density in the fuel to counterbalance the decrease in enrichment. One of the most promising new fuel candidate is U-Mo alloy, in a U-Mo/Al dispersion fuel form, due to its high uranium loading as well as excellent irradiation resistance performance, is being developed extensively to convert from HEU fuel to LEU fuel for high-performance research reactors. However, the formation of an interaction layer (IL) between U-Mo particles and the Al matrix, and the associated pore formation, under high heat flux and high burnup conditions, degrade the irradiation performance of the U-Mo/Al dispersion fuel. From the recent tests results accumulated from the surface engineering of low enriched uranium fuel (SELENIUM) and MIR reactor displayed that a surface barrier coating like physical vapor deposited (PVD) zirconium nitride (ZrN) can significantly reduce the interaction layer. The barrier coating performed well at low burn up but above a fluence rate of 5x 1021 ions/cm2 the swelling reappeared due to formation interaction layer. With this result in mind the objective of this research was to develop an ultrathin ZrN coating over particulate uranium-molybdenum nuclear fuel using a modified savannah 200 atomic layer deposition (ALD) system. This is done in support of the US Department of Energy's (DOE) effort to slow down the interaction at fluence rate and reach higher burn up for high power research reactor. The low-pressure Savannah 200 ALD system is modified to be designed as a batch powder coating system using the

Power ultrasound irradiation during the alkaline etching process of the 2024 aluminum alloy

NASA Astrophysics Data System (ADS)

Moutarlier, V.; Viennet, R.; Rolet, J.; Gigandet, M. P.; Hihn, J. Y.

2015-11-01

Prior to any surface treatment on an aluminum alloy, a surface preparation is necessary. This commonly consists in performing an alkaline etching followed by acid deoxidizing. In this work, the use of power ultrasound irradiation during the etching step on the 2024 aluminum alloy was studied. The etching rate was estimated by weight loss, and the alkaline film formed during the etching step was characterized by glow discharge optical emission spectrometry (GDOES) and scanning electron microscope (SEM). The benefit of power ultrasound during the etching step was confirmed by pitting potential measurement in NaCl solution after a post-treatment (anodizing).

Crevice Corrosion Behavior of 45 Molybdenum-Containing Stainless Steels in Seawater.

DTIC Science & Technology

1981-12-01

Armco, Avesta Jernverks, Cabot, Carpenter Technology, Crucible, Eastern, Firth-Brown, Huntington, Jessup, Langley Alloys, and Uddeholm. 16...Department of Energy, Report ANL/OTEC-BCM-022. 7. Wallen, B., and M. Liljas, " Avesta 254 SMO - A New, High Molybdenum Stainless Steel," presented at NKM8...1977).; 11. Wallen, B., " Avesta 254 SMO - A Stainless Steel for Seawater Service," presented at the Advanced Stainless Steels for Turbine Condensors

DART model for irradiation-induced swelling of uranium silicide dispersion fuel elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J.; Hofman, G.L.

1999-04-01

Models for the interaction of uranium silicide dispersion fuels with an aluminum matrix, for the resultant reaction product swelling, and for the calculation of the stress gradient within the fuel particles are described within the context of DART fission-gas-induced swelling models. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by comparing DART calculations with irradiation data for the swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al in variously designed dispersion fuel elements.

The effect of low energy helium ion irradiation on tungsten-tantalum (W-Ta) alloys under fusion relevant conditions

NASA Astrophysics Data System (ADS)

Gonderman, S.; Tripathi, J. K.; Novakowski, T. J.; Sizyuk, T.; Hassanein, A.

2017-08-01

Currently, tungsten remains the best candidate for plasma-facing components (PFCs) for future fusion devices because of its high melting point, low erosion, and strong mechanical properties. However, continued investigation has shown tungsten to undergo severe morphology changes under fusion-like conditions. These results motivate the study of innovative PFC materials which are resistant to surface morphology evolution. The goal of this work is to examine tungsten-tantalum (W-Ta) alloys, a potential PFC material, and their response to low energy helium ion irradiation. Specifically, W-Ta samples are exposed to 100 eV helium irradiations with a flux of 1.15 × 1021 ions m-2 s-1, at 873 K, 1023 K, and 1173 K for 1 h duration. Scanning electron microscopy (SEM) reveals significant changes in surface deterioration due to helium ion irradiation as a function of both temperature and tantalum concentration in W-Ta samples. X-Ray Diffraction (XRD) studies show a slight lattice parameter expansion in W-Ta alloy samples compared to pure W samples. The observed lattice parameter expansion in W-Ta alloy samples (proportional to increasing Ta wt.% concentrations) reflect significant differences observed in the evolution of surface morphology, i.e., fuzz development processes for both increasing Ta wt.% concentration and target temperature. These results suggest a correlation between the observed morphology differences and the induced crystal structure change caused by the presence of tantalum. Shifts in the XRD peaks before and after 100 eV helium irradiation with a flux of 1.15 × 1021 ions m-2 s-1, 1023 K, for 1 h showed a significant difference in the magnitude of the shift. This has suggested a possible link between the atomic spacing of the material and the accumulated damage. Ongoing research is needed on W-Ta alloys and other innovative materials for their application as irradiation resistant materials in future fusion or irradiation environments.

Nonequilibrium segregation and phase instability in alloy films during elevated-temperature irradiation in a high-voltage electron microscope

NASA Astrophysics Data System (ADS)

Lam, N. Q.; Okamoto, P. R.

1984-05-01

The effects of defect-production rate gradients, caused by the radial nonuniformity in the electron flux distribution, on solute segregation and phase stability in alloy films undergoing high-voltage electron-microscope (HVEM) irradiation at high temperatures are assessed. Two-dimensional (axially symmetric) compositional redistributions were calculated, taking into account both axial and transverse radial defect fluxes. It was found that when highly focused beams were employed radiation-induced segregation consisted of two stages: dominant axial segregation at the film surfaces at short irradiation times and competitive radial segregation at longer times. The average alloy composition within the irradiated region could differ greatly from that irradiated with a uniform beam, because of the additional atom transport from or to the region surrounding the irradiated zone under the influence of radial fluxes. Damage-rate gradient effects must be taken into account when interpreting in-situ HVEM observations of segregation-induced phase instabilities. The theoretical predictions are compared with experimental observations of the temporal and spatial dependence of segregation-induced precipitation in thin films of Ni-Al, Ni-Ge and Ni-Si solid solutions.

Characteristics of surface modified Ti-6Al-4V alloy by a series of YAG laser irradiation

NASA Astrophysics Data System (ADS)

Zeng, Xian; Wang, Wenqin; Yamaguchi, Tomiko; Nishio, Kazumasa

2018-01-01

In this study, a double-layer Ti (C, N) film was successfully prepared on Ti-6Al-4V alloy by a series of YAG laser irradiation in nitrogen atmosphere, aiming at improving the wear resistance. The effects of laser irradiation pass upon surface chemical composition, microstructures and hardness were investigated. The results showed that the surface chemicals were independent from laser irradiation pass, which the up layer of film was a mixture of TiN and TiC0.3N0.7, and the down layer was nitrogen-rich α-Ti. Both the surface roughness and hardness increased as raising the irradiation passes. However, surface deformation and cracks happened in the case above 3 passes' irradiation. The wear resistance of laser modified sample by 3 passes was improved approximately by 37 times compared to the as received substrate. Moreover, the cytotoxic V ion released from laser modified sample was less than that of as received Ti-6Al-4V alloy in SBF, suggesting the potentiality of a new try to modify the sliding part of Ti-based hard tissue implants in future biomedical application.

Redox-Mediated Stabilization in Zinc Molybdenum Nitrides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arca, Elisabetta; Lany, Stephan; Perkins, John D.

We report on the theoretical prediction and experimental realization of new ternary zinc molybdenum nitride compounds. We used theory to identify previously unknown ternary compounds in the Zn-Mo-N systems, Zn 3MoN 4 and ZnMoN 2, and to analyze their bonding environment. Experiments show that Zn-Mo-N alloys can form in broad composition range from Zn 3MoN 4 to ZnMoN 2 in the wurtzite-derived structure, accommodating very large off-stoichiometry. Interestingly, the measured wurtzite-derived structure of the alloys is metastable for the ZnMoN 2 stoichiometry, in contrast to the Zn 3MoN 4 stoichiometry, where ordered wurtzite is predicted to be the ground state.more » The formation of Zn 3MoN 4-ZnMoN 2 alloy with wurtzite-derived crystal structure is enabled by the concomitant ability of Mo to change oxidation state from +VI in Zn 3MoN 4 to +IV in ZnMoN 2, and the capability of Zn to contribute to the bonding states of both compounds, an effect that we define as 'redox-mediated stabilization.' The stabilization of Mo in both the +VI and +IV oxidation states is due to the intermediate electronegativity of Zn, which enables significant polar covalent bonding in both Zn 3MoN 4 and ZnMoN 2 compounds. The smooth change in the Mo oxidation state between Zn 3MoN 4 and ZnMoN 2 stoichiometries leads to a continuous change in optoelectronic properties - from resistive and semitransparent Zn 3MoN 4 to conductive and absorptive ZnMoN 2. The reported redox-mediated stabilization in zinc molybdenum nitrides suggests there might be many undiscovered ternary compounds with one metal having an intermediate electronegativity, enabling significant covalent bonding, and another metal capable of accommodating multiple oxidation states, enabling stoichiometric flexibility.« less

Redox-Mediated Stabilization in Zinc Molybdenum Nitrides

DOE PAGES

Arca, Elisabetta; Lany, Stephan; Perkins, John D.; ...

2018-03-01

We report on the theoretical prediction and experimental realization of new ternary zinc molybdenum nitride compounds. We used theory to identify previously unknown ternary compounds in the Zn-Mo-N systems, Zn 3MoN 4 and ZnMoN 2, and to analyze their bonding environment. Experiments show that Zn-Mo-N alloys can form in broad composition range from Zn 3MoN 4 to ZnMoN 2 in the wurtzite-derived structure, accommodating very large off-stoichiometry. Interestingly, the measured wurtzite-derived structure of the alloys is metastable for the ZnMoN 2 stoichiometry, in contrast to the Zn 3MoN 4 stoichiometry, where ordered wurtzite is predicted to be the ground state.more » The formation of Zn 3MoN 4-ZnMoN 2 alloy with wurtzite-derived crystal structure is enabled by the concomitant ability of Mo to change oxidation state from +VI in Zn 3MoN 4 to +IV in ZnMoN 2, and the capability of Zn to contribute to the bonding states of both compounds, an effect that we define as 'redox-mediated stabilization.' The stabilization of Mo in both the +VI and +IV oxidation states is due to the intermediate electronegativity of Zn, which enables significant polar covalent bonding in both Zn 3MoN 4 and ZnMoN 2 compounds. The smooth change in the Mo oxidation state between Zn 3MoN 4 and ZnMoN 2 stoichiometries leads to a continuous change in optoelectronic properties - from resistive and semitransparent Zn 3MoN 4 to conductive and absorptive ZnMoN 2. The reported redox-mediated stabilization in zinc molybdenum nitrides suggests there might be many undiscovered ternary compounds with one metal having an intermediate electronegativity, enabling significant covalent bonding, and another metal capable of accommodating multiple oxidation states, enabling stoichiometric flexibility.« less

TUNGSTEN INTERFERENCE IN VOLUMETRIC ANALYSIS OF URANIUM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dufour, R.F.; Articolo, O.

1958-08-01

Tungsten was found to have a negligible effect on the determination of uranium in uranium-zirconium alloys by the Jones reductor-dichromate method used at KAPL. The tungstate ion interferred seriously and gave high results. However, the soluble tungsten was precipitated by intensive fuming with sulfuric acid and rendered ineffective in tbe subsequent oxidationreduction reactions of the uranium. (auth)

Nickel aluminide alloy suitable for structural applications

DOEpatents

Liu, C.T.

1998-03-10

Alloys are disclosed for use in structural applications based upon NiAl to which are added selected elements to enhance room temperature ductility and high temperature strength. Specifically, small additions of molybdenum produce a beneficial alloy, while further additions of boron, carbon, iron, niobium, tantalum, zirconium and hafnium further improve performance of alloys at both room temperature and high temperatures. A preferred alloy system composition is Ni--(49.1{+-}0.8%)Al--(1.0{+-}0.8%)Mo--(0.7 + 0.5%)Nb/Ta/Zr/Hf--(nearly zero to 0.03%)B/C, where the % is at. % in each of the concentrations. All alloys demonstrated good oxidation resistance at the elevated temperatures. The alloys can be fabricated into components using conventional techniques. 4 figs.

Effect of laser irradiation conditions on the laser welding strength of cobalt-chromium and gold alloys.

PubMed

Kikuchi, Hisaji; Kurotani, Tomoko; Kaketani, Masahiro; Hiraguchi, Hisako; Hirose, Hideharu; Yoneyama, Takayuki

2011-09-01

Using tensile tests, this study investigated differences in the welding strength of casts of cobalt-chromium and gold alloys resulting from changes in the voltage and pulse duration in order to clarify the optimum conditions of laser irradiation for achieving favorable welding strength. Laser irradiation was performed at voltages of 150 V and 170 V with pulse durations of 4, 8, and 12 ms. For cobalt-chromium and gold alloys, it was found that a good welding strength could be achieved using a voltage of 170 V, a pulse duration of 8 ms, and a spot diameter of 0.5 mm. However, when the power density was set higher than this, defects tended to occur, suggesting the need for care when establishing welding conditions.

Modification of surface properties of copper-refractory metal alloys

DOEpatents

Verhoeven, John D.; Gibson, Edwin D.

1993-10-12

The surface properties of copper-refractory metal (CU-RF) alloy bodies are modified by heat treatments which cause the refractory metal to form a coating on the exterior surfaces of the alloy body. The alloys have a copper matrix with particles or dendrites of the refractory metal dispersed therein, which may be niobium, vanadium, tantalum, chromium, molybdenum, or tungsten. The surface properties of the bodies are changed from those of copper to that of the refractory metal.

TEM in situ micropillar compression tests of ion irradiated oxide dispersion strengthened alloy

NASA Astrophysics Data System (ADS)

Yano, K. H.; Swenson, M. J.; Wu, Y.; Wharry, J. P.

2017-01-01

The growing role of charged particle irradiation in the evaluation of nuclear reactor candidate materials requires the development of novel methods to assess mechanical properties in near-surface irradiation damage layers just a few micrometers thick. In situ transmission electron microscopic (TEM) mechanical testing is one such promising method. In this work, microcompression pillars are fabricated from a Fe2+ ion irradiated bulk specimen of a model Fe-9%Cr oxide dispersion strengthened (ODS) alloy. Yield strengths measured directly from TEM in situ compression tests are within expected values, and are consistent with predictions based on the irradiated microstructure. Measured elastic modulus values, once adjusted for the amount of deformation and deflection in the base material, are also within the expected range. A pillar size effect is only observed in samples with minimum dimension ≤100 nm due to the low inter-obstacle spacing in the as received and irradiated material. TEM in situ micropillar compression tests hold great promise for quantitatively determining mechanical properties of shallow ion-irradiated layers.

REGENERATION OF FISSION-PRODUCT-CONTAINING MAGNESIUM-THORIUM ALLOYS

DOEpatents

Chiotti, P.

1964-02-01

A process of regenerating a magnesium-thorium alloy contaminated with fission products, protactinium, and uranium is presented. A molten mixture of KCl--LiCl-MgCl/sub 2/ is added to the molten alloy whereby the alkali, alkaline parth, and rare earth fission products (including yttrium) and some of the thorium and uranium are chlorinated and

Molybdenum carbide supported nickel-molybdenum alloys for synthesis gas production via partial oxidation of surrogate biodiesel

NASA Astrophysics Data System (ADS)

Shah, Shreya; Marin-Flores, Oscar G.; Norton, M. Grant; Ha, Su

2015-10-01

In this study, NiMo alloys supported on Mo2C are synthesized by wet impregnation for partial oxidation of methyl oleate, a surrogate biodiesel, to produce syngas. When compared to single phase Mo2C, the H2 yield increases from 70% up to >95% at the carbon conversion of ∼100% for NiMo alloy nanoparticles that are dispersed over the Mo2C surface. Supported NiMo alloy samples are prepared at two different calcination temperatures in order to determine its effect on particle dispersion, crystalline phase and catalytic properties. The reforming test data indicate that catalyst prepared at lower calcination temperature shows better nanoparticle dispersion over the Mo2C surface, which leads to higher initial performance when compared to catalysts synthesized at higher calcination temperature. Activity tests using the supported NiMo alloy on Mo2C that are calcined at the lower temperature of 400 °C shows 100% carbon conversion with 90% H2 yield without deactivation due to coking over 24 h time-on-stream.

Point defect evolution in Ni, NiFe and NiCr alloys from atomistic simulations and irradiation experiments

DOE PAGES

Aidhy, Dilpuneet S.; Lu, Chenyang; Jin, Ke; ...

2015-08-08

Using molecular dynamics simulations, we elucidate irradiation-induced point defect evolution in fcc pure Ni, Ni 0.5Fe 0.5, and Ni 0.8Cr 0.2 solid solution alloys. We find that irradiation-induced interstitials form dislocation loops that are of 1/3 <111>{111}-type, consistent with our experimental results. While the loops are formed in all the three materials, the kinetics of formation is considerably slower in NiFe and NiCr than in pure Ni, indicating that defect migration barriers and extended defect formation energies could be higher in the alloys than pure Ni. As a result, while larger size clusters are formed in pure Ni, smaller andmore » more clusters are observed in the alloys. The vacancy diffusion occurs at relatively higher temperatures than interstitials, and their clustering leads to formation of stacking fault tetrahedra, also consistent with our experiments. The results also show that the surviving Frenkel pairs are composition-dependent and are largely Ni dominated.« less

Helium sequestration at nanoparticle-matrix interfaces in helium + heavy ion irradiated nanostructured ferritic alloys

DOE PAGES

Parish, Chad M.; Unocic, Kinga A.; Tan, Lizhen; ...

2016-10-24

Here we irradiated four ferritic alloys with energetic Fe and He ions: one castable nanostructured alloy (CNA) containing Ti-W-Ta-carbides, and three nanostructured ferritic alloys (NFAs). The NFAs were: 9Cr containing Y-Ti-O nanoclusters, and two Fe-12Cr-5Al NFAs containing Y-Zr-O or Y-Hf-O clusters. All four were subjected to simultaneous dual-beam Fe + He ion implantation (650 °C, ~50 dpa, ~15 appm He/dpa), simulating fusion-reactor conditions. Examination using scanning/transmission electron microscopy (STEM) revealed high-number-density helium bubbles of ~8 nm, ~10 21 m -3 (CNA), and of ~3 nm, 10 23 m -3 (NFAs). STEM combined with multivariate statistical analysis data mining suggests thatmore » the precipitate-matrix interfaces in all alloys survived ~50 dpa at 650 °C and serve as effective helium trapping sites. All alloys appear viable structural material candidates for fusion or advanced fission energy systems. Finally, among these developmental alloys the NFAs appear to sequester the helium into smaller bubbles and away from the grain boundaries more effectively than the early-generation CNA.« less

A combined APT and SANS investigation of α' phase precipitation in neutron-irradiated model FeCrAl alloys

DOE PAGES

Briggs, Samuel A.; Edmondson, Philip D.; Littrell, Kenneth C.; ...

2017-03-01

Here, FeCrAl alloys are currently under consideration for accident-tolerant fuel cladding applications in light water reactors owing to their superior high-temperature oxidation and corrosion resistance compared to the Zr-based alloys currently employed. However, their performance could be limited by precipitation of a Cr-rich α' phase that tends to embrittle high-Cr ferritic Fe-based alloys. In this study, four FeCrAl model alloys with 10–18 at.% Cr and 5.8–9.3 at.% Al were neutron-irradiated to nominal damage doses up to 7.0 displacements per atom at a target temperature of 320 °C. Small angle neutron scattering techniques were coupled with atom probe tomography to assessmore » the composition and morphology of the resulting α' precipitates. It was demonstrated that Al additions partially destabilize the α' phase, generally resulting in precipitates with lower Cr contents when compared with binary Fe-Cr systems. The precipitate morphology evolution with dose exhibited a transient coarsening regime akin to previously observed behavior in aged Fe-Cr alloys. Similar behavior to predictions of the LSW/UOKV models suggests that α' precipitation in irradiated FeCrAl is a diffusion-limited process with coarsening mechanisms similar to those in thermally aged high-Cr ferritic alloys.« less

The Development of the Low-Cost Titanium Alloy Containing Cr and Mn Alloying Elements

NASA Astrophysics Data System (ADS)

Zhu, Kailiang; Gui, Na; Jiang, Tao; Zhu, Ming; Lu, Xionggang; Zhang, Jieyu; Li, Chonghe

2014-04-01

The α + β-type Ti-4.5Al-6.9Cr-2.3Mn alloy has been theoretically designed on the basis of assessment of the Ti-Al-Cr-Mn thermodynamic system and the relationship between the molybdenum equivalent and mechanical properties of titanium alloys. The alloy is successfully prepared by the split water-cooled copper crucible, and its microstructures and mechanical properties at room temperature are investigated using the OM, SEM, and the universal testing machine. The results show that the Ti-4.5Al-6.9Cr-2.3Mn alloy is an α + β-type alloy which is consistent with the expectation, and its fracture strength, yield strength, and elongation reach 1191.3, 928.4 MPa, and 10.7 pct, respectively. Although there is no strong segregation of alloying elements under the condition of as-cast, the segregation of Cr and Mn is obvious at the grain boundary after thermomechanical treatment.

THE ACCURATE DETERMINATION OF MICROGRAM AMOUNTS OF BORON IN ALUMINUM AND ALUMINUM-URANIUM ALLOYS BY THE METHYL BORATE-CURCUMIN-OXALIC ACID METHOD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crocker, I.H.

1958-10-01

A method was developed for the deternninntion of boron in aluminum and aluminum--uranium alloys in which the boron concentration is 30 ppm or more. Boron is separated by distillation as methyl borate from a hydrochloric acid solution of the alloy and is determined spectrophotometrically by the boric acid-- curcumin-oxalic acid color reaction. A precision of plus or minus 2% is attain able when the determination is penformed with the utmost care. The accuracy is such that no bias need be given when a calibration curve is used. (auth)

Characterization of weld metal microstructure in a Ni-30Cr alloy with additions of niobium and molybdenum

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheeling, Rebecca A., E-mail: wheeling.8@osu.edu; Lippold, John C., E-mail: lippold.1@osu.edu

2016-05-15

Additions of niobium (Nb) and molybdenum (Mo) were made to an Alloy 690 base alloy in order to investigate the formation of a eutectic constituent at the end of solidification and to evaluate the effect of the eutectic liquid on backfilling (or healing) of solidification cracks. Solidification cracking was induced using the cast pin tear test (CPTT) and regions of backfilling were located and characterized via optical and electron microscopy. Computational predictions of fraction eutectic and composition of the eutectic constituent were compared to experimental findings and were found to correlate well in both cases. The extent of crack backfillingmore » increased significantly with increasing Nb content, but the addition of Mo did not seem to influence the amount of eutectic constituent or the degree of backfilling. SEM/EDS analysis confirmed that the eutectic composition is constant and that increasing Nb above 4 wt% has little effect on expanding the solidification temperature range, but has a beneficial effect on mitigating solidification cracking by a crack healing effect. - Highlights: • Increasing fraction eutectic as a function of Nb, as predicted by ThermoCalc™, is consistent with image analysis results. • Nb, unlike Mo, had a significant effect on the fraction eutectic formed. • Both influence the composition of the eutectic. • Thermocalc™ predictions regarding Nb content in eutectic are consistent with EDS results, but are high for the Mo content. • Increased levels of niobium resulted in a higher degree of crack backfilling and leads to a lower cracking susceptibility. • Mo may influence the eutectic liquid along solidification grain boundaries, improving backfill and thus cracking resistance.« less

Dislocation loop evolution during in-situ ion irradiation of model FeCrAl alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haley, Jack C.; Briggs, Samuel A.; Edmondson, Philip D.

Model FeCrAl alloys of Fe-10%Cr-5%Al, Fe-12%Cr-4.5%Al, Fe-15%Cr-4%Al, and Fe-18%Cr-3%Al (in wt %) were irradiated with 1 MeV Kr++ ions in-situ with transmission electron microscopy to a dose of 2.5 displacements per atom (dpa) at 320 °C. In all cases, the microstructural damage consisted of dislocation loops with ½< 111 > and <100 > Burgers vectors. The proportion of ½< 111 > dislocation loops varied from ~50% in the Fe-10%Cr-5%Al model alloy and the Fe-18Cr%-3%Al model alloy to a peak of ~80% in the model Fe-15%Cr-4.5%Al alloy. The dislocation loop volume density increased with dose for all alloys and showed signsmore » of approaching an upper limit. The total loop populations at 2.5 dpa had a slight (and possibly insignificant) decline as the chromium content was increased from 10 to 15 wt %, but the Fe-18%Cr-3%Al alloy had a dislocation loop population ~50% smaller than the other model alloys. As a result, the largest dislocation loops in each alloy had image sizes of close to 20 nm in the micrographs, and the median diameters for all alloys ranged from 6 to 8 nm. Nature analysis by the inside-outside method indicated most dislocation loops were interstitial type.« less

Dislocation loop evolution during in-situ ion irradiation of model FeCrAl alloys

DOE PAGES

Haley, Jack C.; Briggs, Samuel A.; Edmondson, Philip D.; ...

2017-07-06

Model FeCrAl alloys of Fe-10%Cr-5%Al, Fe-12%Cr-4.5%Al, Fe-15%Cr-4%Al, and Fe-18%Cr-3%Al (in wt %) were irradiated with 1 MeV Kr++ ions in-situ with transmission electron microscopy to a dose of 2.5 displacements per atom (dpa) at 320 °C. In all cases, the microstructural damage consisted of dislocation loops with ½< 111 > and <100 > Burgers vectors. The proportion of ½< 111 > dislocation loops varied from ~50% in the Fe-10%Cr-5%Al model alloy and the Fe-18Cr%-3%Al model alloy to a peak of ~80% in the model Fe-15%Cr-4.5%Al alloy. The dislocation loop volume density increased with dose for all alloys and showed signsmore » of approaching an upper limit. The total loop populations at 2.5 dpa had a slight (and possibly insignificant) decline as the chromium content was increased from 10 to 15 wt %, but the Fe-18%Cr-3%Al alloy had a dislocation loop population ~50% smaller than the other model alloys. As a result, the largest dislocation loops in each alloy had image sizes of close to 20 nm in the micrographs, and the median diameters for all alloys ranged from 6 to 8 nm. Nature analysis by the inside-outside method indicated most dislocation loops were interstitial type.« less

Physicochemical investigation of NiAl with small molybdenum additions

NASA Technical Reports Server (NTRS)

Troshkina, V. A.; Kucherenko, L. A.; Fadeeva, V. I.; Aristova, N. M.

1982-01-01

Specimens of four cast NiAl alloys, three of them containing 0.5, 1.0 and 1.5 at. % Mo., were homogenized for 10, 10, and 140 hr at 1373, 1523 and 1273 K, respectively, then kept at 1073, 1173 and 1323 K for 60, 120 and 3 hr, respectively, and quenched in icy water. The precipitation of a metastable Ni3Mo phase was observed at temperatures between 1073 and 1523 K. Molybdenum substituted for nickel was found to inhibit the lattice disordering in NiAl at 1073 and 1523 K.

Application of a three-feature dispersed-barrier hardening model to neutron-irradiated Fe-Cr model alloys

NASA Astrophysics Data System (ADS)

Bergner, F.; Pareige, C.; Hernández-Mayoral, M.; Malerba, L.; Heintze, C.

2014-05-01

An attempt is made to quantify the contributions of different types of defect-solute clusters to the total irradiation-induced yield stress increase in neutron-irradiated (300 °C, 0.6 dpa), industrial-purity Fe-Cr model alloys (target Cr contents of 2.5, 5, 9 and 12 at.% Cr). Former work based on the application of transmission electron microscopy, atom probe tomography, and small-angle neutron scattering revealed the formation of dislocation loops, NiSiPCr-enriched clusters and α‧-phase particles, which act as obstacles to dislocation glide. The values of the dimensionless obstacle strength are estimated in the framework of a three-feature dispersed-barrier hardening model. Special attention is paid to the effect of measuring errors, experimental details and model details on the estimates. The three families of obstacles and the hardening model are well capable of reproducing the observed yield stress increase as a function of Cr content, suggesting that the nanostructural features identified experimentally are the main, if not the only, causes of irradiation hardening in these model alloys.

Stability of nanoclusters in an oxide dispersion strengthened alloy under neutron irradiation

DOE PAGES

Liu, Xiang; Miao, Yinbin; Wu, Yaqiao; ...

2017-06-01

In this paper, 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. Finally, the chemical modifications indicate the equilibrium size is indeed a balance of two competing processes: radiationmore » enhanced diffusion and collisional dissolution.« less

URANIUM RECOVERY PROCESS

DOEpatents

Hyman, H.H.; Dreher, J.L.

1959-07-01

The recovery of uranium from the acidic aqueous metal waste solutions resulting from the bismuth phosphate carrier precipitation of plutonium from solutions of neutron irradiated uranium is described. The waste solutions consist of phosphoric acid, sulfuric acid, and uranium as a uranyl salt, together with salts of the fission products normally associated with neutron irradiated uranium. Generally, the process of the invention involves the partial neutralization of the waste solution with sodium hydroxide, followed by conversion of the solution to a pH 11 by mixing therewith sufficient sodium carbonate. The resultant carbonate-complexed waste is contacted with a titanated silica gel and the adsorbent separated from the aqueous medium. The aqueous solution is then mixed with sufficient acetic acid to bring the pH of the aqueous medium to between 4 and 5, whereby sodium uranyl acetate is precipitated. The precipitate is dissolved in nitric acid and the resulting solution preferably provided with salting out agents. Uranyl nitrate is recovered from the solution by extraction with an ether such as diethyl ether.

HEAT TREATED U-Nb ALLOYS

DOEpatents

McGeary, R.K.; Justusson, W.M.

1959-11-24

A fuel element for a nuclear reactor is described comprising an alloy containing uranium and from 7 to 20 wt.% niobium, the alloy being substantially in the gamma phase and having been produced by working an ingot of the alloy into the desired shape, homogenizing it by annealing it at a temperature in the gamma phase field, and quenching it to retain the gamma phase structure of the alloy.

In situ resistivity measurements of RAFM base alloys at cryogenic temperatures: The effect of proton irradiation

NASA Astrophysics Data System (ADS)

Gómez-Ferrer, B.; Vila, R.; Jiménez-Rey, D.; Ortiz, C. J.; Mota, F.; García, J. M.; Rodríguez, A.

2014-04-01

A four-probe technique for measurement of electrical resistance on low-temperature ion-irradiated metallic sheets is described. The design, temperature control system, preparation method of samples and the resistivity measurements are described in detail. The resistivity recovery (RR) curve has been measured on a Fe-5%Cr model alloy irradiated with 5 MeV protons. The procedure to obtain the RR derivative curve is outlined and experimental errors are identified and quantified. Special care has been taken to use a sample with very low impurity content and low dislocation density (1.2 × 108 cm-2). Thus, effects in recovery spectrum of the Fe-5%Cr alloy are only due to the presence of Cr and irradiation defects, which will be mainly Frenkel Pairs (FPs) given that the mean energy of the Primary Knock-on Atoms (PKA) is close to 0.35 keV. The results obtained for the Fe-5%Cr under 5 MeV proton irradiation are found to be in overall agreement with previous experimental measurements performed under electron irradiation although some differences appear probably due to the different spatial distribution of the created defects and the higher temperature resolution of annealing steps. The RR spectrum obtained reveals the appearance of the structure of stages I and II and also a partial suppression of the stage III peak with respect to previous results obtained after electron irradiation. The stage III suppression is explained as a superposition of vacancy recombination effects and short-range ordering (SRO) effects which are apparently dependent on the spatial distribution of defects created during irradiation. Moreover, recombination phenomena are observed beyond stage III up to 500 K.

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montierth, Leland M.

2016-07-19

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element designmore » for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.« less

Process for massively hydriding zirconium--uranium fuel elements

DOEpatents

Katz, N.H.

1973-12-01

A method is described of hydriding uranium-zirconium alloy by heating the alloy in a vacuum, introducing hydrogen and maintaining an elevated temperature until occurrence of the beta--delta phase transformation and isobarically cooling the composition. (Official Gazette)

Molybdenum Trafficking for Nitrogen Fixation†

PubMed Central

Hernandez, Jose A.; George, Simon J.; Rubio, Luis M.

2009-01-01

The molybdenum nitrogenase is responsible for most biological nitrogen fixation, a prokaryotic metabolic process that determines the global biogeochemical cycles of nitrogen and carbon. Here we describe the trafficking of molybdenum for nitrogen fixation in the model diazotrophic bacterium Azotobacter vinelandii. The genes and proteins involved in molybdenum uptake, homeostasis, storage, regulation, and nitrogenase cofactor biosynthesis are reviewed. Molybdenum biochemistry in A. vinelandii reveals unexpected mechanisms and a new role for iron-sulfur clusters in the sequestration and delivery of molybdenum. PMID:19772354

Effect of ternary addition and gamma-irradiation on the characteristics of rapidly solidified Pb-base alloys

NASA Astrophysics Data System (ADS)

Abd El-Khalek, A. M.

The properties of a series of rapidly solidified Pb-Sb-3-Sn-x alloys ( x =0-2.5 wt.%) irradiated with gamma-rays were studied. Variations in the internal friction, Q(-1) , thermal diffusivity D th and dynamic Young's modulus Y were traced before and after irradiation by applying the resonance technique. Variations of specific heat C-p were obtained from DTA thermograms. Structure parameters were obtained from the X-rays diffraction patterns. A marked change in the behaviour of the measured parameters was observed at 1.5 wt.% Sn addition. Besides, irradiation induced defects increased the level of the measured hardening parameters.

Experimental investigations on thermo mechanical behaviour of aluminium alloys subjected to tensile loading and laser irradiation

NASA Astrophysics Data System (ADS)

Jelani, Mohsan; Li, Zewen; Shen, Zhonghua; Sardar, Maryam; Tabassum, Aasma

2017-05-01

The present work reports the investigation of the thermal and mechanical behaviour of aluminium alloys under the combined action of tensile loading and laser irradiations. The two types of aluminium alloys (Al-1060 and Al-6061) are used for the experiments. The continuous wave Ytterbium fibre laser (wavelength 1080 nm) was employed as irradiation source, while tensile loading was provided by tensile testing machine. The effects of various pre-loading and laser power densities on the failure time, temperature distribution and on deformation behaviour of aluminium alloys are analysed. The experimental results represents the significant reduction in failure time and temperature for higher laser powers and for high load values, which implies that preloading may contribute a significant role in the failure of the material at elevated temperature. The reason and characterization of material failure by tensile and laser loading are explored in detail. A comparative behaviour of under tested materials is also investigated. This work suggests that, studies considering only combined loading are not enough to fully understand the mechanical behaviour of under tested materials. For complete characterization, one must consider the effect of heating as well as loading rate.

Creating poly(ethylene glycol) film on the surface of NiTi alloy by gamma irradiation

NASA Astrophysics Data System (ADS)

Yu, Hongyan; Yan, Jin; Ma, Huiling; Zeng, Xinmiao; Liu, Yang; Zhao, Xinqing

2015-07-01

NiTi alloy has been extensively utilized as biomaterials owing to its unique shape memory effect, superelasticity and biocompatibility. However, concern with the toxic and allergic responses of nickel potentially releasing from implants stimulated lots of researches of modification on NiTi alloy surface. Creating chemical bond attachment of bioorganic film on NiTi alloy surface could effectively inhibit Ni releasing and obtain bioactive functions for further application. In this work, to get a bioorganic surface, NiTi alloy was modified with poly(ethylene glycol) (PEG) film by gamma ray induced grafting or crosslinking. X-ray diffraction (XRD) spectrum, water contact angle geometer and X-ray photoelectron spectroscopy (XPS) techniques were used to characterize the NiTi surface. The results indicated that PEG was covalent bonded on NiTi alloy surface. Fluorescence microscope (FM) images for morphology of 1 day osteoblast culture on the PEG coated NiTi surface showed that PEG could improve cell proliferation on NiTi surface. Our work offers a way to introduce a bioorganic metal surface by gamma irradiation.

Metal alloy coatings and methods for applying

DOEpatents

Merz, Martin D.; Knoll, Robert W.

1991-01-01

A method of coating a substrate comprises plasma spraying a prealloyed feed powder onto a substrate, where the prealloyed feed powder comprises a significant amount of an alloy of stainless steel and at least one refractory element selected from the group consisting of titanium, zirconium, hafnium, niobium, tantalum, molybdenum, and tungsten. The plasma spraying of such a feed powder is conducted in an oxygen containing atmosphere and forms an adherent, corrosion resistant, and substantially homogenous metallic refractory alloy coating on the substrate.

The Nature of Surface Oxides on Corrosion-Resistant Nickel Alloy Covered by Alkaline Water

PubMed Central

2010-01-01

A nickel alloy with high chrome and molybdenum content was found to form a highly resistive and passive oxide layer. The donor density and mobility of ions in the oxide layer has been determined as a function of the electrical potential when alkaline water layers are on the alloy surface in order to account for the relative inertness of the nickel alloy in corrosive environments. PMID:20672134

Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Y.; Chopra, O. K.; Soppet, W. K.

2010-02-16

Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier testsmore » with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.« less

Molybdenum cell for x-ray diffraction measurements of fluid alkali metals at high temperatures and high pressures

NASA Astrophysics Data System (ADS)

Matsuda, Kazuhiro; Tamura, Kozaburo; Katoh, Masahiro; Inui, Masanori

2004-03-01

We have developed a sample cell for x-ray diffraction measurements of fluid alkali metals at high temperatures and high pressures. All parts of the cell are made of molybdenum which is resistant to the chemical corrosion of alkali metals. Single crystalline molybdenum disks electrolytically thinned down to 40 μm were used as the walls of the cell through which x rays pass. The crystal orientation of the disks was controlled in order to reduce the background from the cell. All parts of the cell were assembled and brazed together using a high-temperature Ru-Mo alloy. Energy dispersive x-ray diffraction measurements have been successfully carried out for fluid rubidium up to 1973 K and 16.2 MPa. The obtained S(Q) demonstrates the applicability of the molybdenum cell to x-ray diffraction measurements of fluid alkali metals at high temperatures and high pressures.

Obtaining and Mechanical Properties of Ti-Mo-Zr-Ta Alloys

NASA Astrophysics Data System (ADS)

Bălţatu, M. S.; Vizureanu, P.; Geantă, V.; Nejneru, C.; Țugui, C. A.; Focşăneanu, S. C.

2017-06-01

Ti-based alloys are successfully used in the area of orthopedic biomaterials for their enhanced biocompatibility, good corrosion and mechanical properties. The most suitable metals as an alloying element for orthopedic biomaterials are zirconium, molybdenum and tantalum because are non toxic and have good properties. The paper purpose development of two alloys of Ti-Mo-Zr-Ta (TMZT) prepared by arc-melting with several mechanical properties determined by microindentation. The mechanical properties analyzed was Vickers hardness and dynamic elasticity modulus. The investigated alloys presents a low Young’s modulus, an important condition of biomaterials for preventing stress shielding phenomenon.

Hardness behavior of binary and ternary niobium alloys at 77 and 300 K

NASA Technical Reports Server (NTRS)

Stephens, J. R.; Witzke, W. R.

1974-01-01

The effects of alloy additions of zirconium, hafnium, molybdenum, tungsten, rhenium, ruthenium, osmium, rhodium, and iridium on the hardness of niobium was determined. Both binary and ternary alloys were investigated by means of hardness tests at 77 K and 300 K. Results showed that atomic size misfit plays a dominant role in controlling hardness of binary niobium alloys. Alloy softening, which occurred at dilute solute additions, is most likely due to an extrinsic mechanism involving interaction between solute elements and interstitial impurities.

Ultrafast laser-induced reproducible nano-gratings on a molybdenum surface

NASA Astrophysics Data System (ADS)

Dar, Mudasir H.; Saad, Nabil A.; Sahoo, Chakradhar; Naraharisetty, Sri Ram G.; Rao Desai, Narayana

2017-02-01

Wavelength-dependent reproducible nano-gratings were produced on a bulk molybdenum surface upon irradiation with femtosecond laser pulses at near normal incidence in ambient air and water environments. The surface morphology of the irradiated surfaces was characterized by field emission scanning electron microscopy. The ripple spacing was observed to decrease by half when the surface was irradiated with the second harmonic of the fundamental 800 nm radiation. Careful choice of the laser parameters such as fluence, scanning speed, polarization and wavelength were observed to be important for the formation of smooth periodic ripples. The mechanism of formation of polarization-dependent periodic ripples is explained based on the interference model. We also demonstrated the use of a laser direct writing technique for the fabrication of periodic subwavelength structures that have potential applications in photonic devices.

The Permo-Triassic uranium deposits of Gondwanaland

NASA Astrophysics Data System (ADS)

le Roux, J. P.; Toens, P. D.

The world's uranium provinces are time bound and occur in five distinct periods ranging from the Proterozoic to the Recent. One of these periods embraces the time of Gondwana sedimentation and probably is related to the proliferation of land plants from the Devonian on-ward. Decaying vegetal matter produced reducing conditions that enhanced uranium precipitation. The association of uranium with molassic basins adjacent to uplifted granitic and volcanic arcs suggests that lithospheric plate subduction, leading to anatexis of basement rocks and andesitic volcanism, created favorable conditions for uranium mineralization. Uranium occurrences of Gondwana age are of four main types: sandstone-hosted, coal-hosted, pelite-hosted, and vein-type deposits. Sandstone-hosted deposits commonly occur in fluviodeltaic sediments and are related to the presence of organic matter. These deposits commonly are enriched in molybdenum and other base metal sulfides and have been found in South Africa, Zimbabwe, Zambia, Angola, Niger, Madagascar, India, Australia, Argentina, and Brazil. Coalhosted deposits contain large reserves of uranium but are of low grade. In Africa they are mostly within the Permian Ecca Group and its lateral equivalents, as in the Springbok Flats, Limpopo, Botswana, and Tanzania basins. Uraniferous black shales are present in the Gabon and Amazon basins but grades are low. Vein-type uranium is found in Argentina, where it occurs in clustered veins crosscutting sedimentary rocks and quartz porphyries.

PREPARATION OF ACTINIDE-ALUMINUM ALLOYS

DOEpatents

Moore, R.H.

1962-09-01

BS>A process is given for preparing alloys of aluminum with plutonium, uranium, and/or thorium by chlorinating actinide oxide dissolved in molten alkali metal chloride with hydrochloric acid, chlorine, and/or phosgene, adding aluminum metal, and passing air and/or water vapor through the mass. Actinide metal is formed and alloyed with the aluminum. After cooling to solidification, the alloy is separated from the salt. (AEC)

PROCESS FOR SEPARATING URANIUM FISSION PRODUCTS

DOEpatents

Spedding, F.H.; Butler, T.A.; Johns, I.B.

1959-03-10

The removal of fission products such as strontium, barium, cesium, rubidium, or iodine from neutronirradiated uranium is described. Uranium halide or elemental halogen is added to melted irradiated uranium to convert the fission products to either more volatile compositions which vaporize from the melt or to higher melting point compositions which separate as solids.

Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

NASA Technical Reports Server (NTRS)

Collins, J. F.; Debogdan, C. E.; Diianni, D. C.

1973-01-01

A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

Molybdenum-base cermet fuel development

NASA Astrophysics Data System (ADS)

Pilger, James P.; Gurwell, William E.; Moss, Ronald W.; White, George D.; Seifert, David A.

Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. The MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermte. This cermet is to have a high matrix density (greater than or equal to 95 percent) for high strength and high thermal conductance coupled with a high particle (UN) porosity (approximately 25 percent) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous Un microspheres become available. Process development was conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and the vacuum hot press consolidation techniques. This paper summarizes the status of these activities.

In situ observation of defect annihilation in Kr ion-irradiated bulk Fe/amorphous-Fe 2 Zr nanocomposite alloy

DOE PAGES

Yu, K. Y.; Fan, Z.; Chen, Y.; ...

2014-08-26

Enhanced irradiation tolerance in crystalline multilayers has received significant attention lately. However, little is known on the irradiation response of crystal/amorphous nanolayers. We report on in situ Kr ion irradiation studies of a bulk Fe 96Zr 4 nanocomposite alloy. Irradiation resulted in amorphization of Fe 2Zr and formed crystal/amorphous nanolayers. α-Fe layers exhibited drastically lower defect density and size than those in large α-Fe grains. In situ video revealed that mobile dislocation loops in α-Fe layers were confined by the crystal/amorphous interfaces and kept migrating to annihilate other defects. This study provides new insights on the design of irradiation-tolerant crystal/amorphousmore » nanocomposites.« less

Duct and cladding alloy

DOEpatents

Korenko, Michael K.

1983-01-01

An austenitic alloy having good thermal stability and resistance to sodium corrosion at 700.degree. C. consists essentially of 35-45% nickel 7.5-14% chromium 0.8-3.2% molybdenum 0.3-1.0% silicon 0.2-1.0% manganese 0-0.1% zirconium 2.0-3.5% titanium 1.0-2.0% aluminum 0.02-0.1% carbon 0-0.01% boron and the balance iron.

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prabhakaran, Ramprashad; Joshi, Vineet V.; Rhodes, Mark A.

2016-10-01

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prabhakaran, Ramprashad; Joshi, Vineet V.; Rhodes, Mark A.

2016-03-30

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

Results of thermal test of metallic molybdenum disk target and fast-acting valve testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Virgo, M.; Chemerisov, S.; Gromov, R.

2016-12-01

This report describes the irradiation conditions for thermal testing of helium-cooled metallic disk targets that was conducted on March 9, 2016, at the Argonne National Laboratory electron linac. The four disks in this irradiation were pressed and sintered by Oak Ridge National Laboratory from molybdenum metal powder. Two of those disks were instrumented with thermocouples. Also reported are results of testing a fast-acting-valve system, which was designed to protect the accelerator in case of a target-window failure.

Uranium chloride extraction of transuranium elements from LWR fuel

DOEpatents

Miller, William E.; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Pierce, R. Dean

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

Uranium chloride extraction of transuranium elements from LWR fuel

DOEpatents

Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

1992-08-25

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure.

FLUORIDE VOLATILITY PROCESS FOR THE RECOVERY OF URANIUM

DOEpatents

Katz, J.J.; Hyman, H.H.; Sheft, I.

1958-04-15

The separation and recovery of uraniunn from contaminants introduced by neutron irradiation by a halogenation and volatilization method are described. The irradiated uranium is dissolved in bromine trifluoride in the liquid phase. The uranium is converted to the BrF/sub 3/ soluble urmium hexafluoride compound whereas the fluorides of certain contaminating elements are insoluble in liquid BrF/sub 3/, and the reaction rate of the BrF/sub 3/ with certain other solid uranium contamirnnts is sufficiently slower than the reaction rate with uranium that substantial portions of these contaminating elements will remain as solids. These solids are then separated from the solution by a distillation, filtration, or centrifugation step. The uranium hexafluoride is then separated from the balance of the impurities and solvent by one or more distillations.

Fundamental Studies of Irradiation-Induced Modifications in Microstructural Evolution and Mechanical Properties of Advanced Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stubbins, James; Heuser, Brent; Hosemann, Peter

This final technical report summarizes the research performed during October 2014 and December 2017, with a focus on investigating the radiation-induced microstructural and mechanical property modifications in optimized advanced alloys for sodium-cooled fast reactor (SFR) structural applications. To accomplish these objectives, the radiation responses of several different advanced alloys, including austenitic steel Alloy 709 (A709) and 316H, and ferritic/ martensitic Fe–9Cr steels T91 and G92, were investigated using a combination of microstructure characterizations and nanoindentation measurements. Different types of irradiation, including ex situ bulk ion irradiation and in situ transmission electron microscopy (TEM) ion irradiation, were employed in this study.more » Radiation-induced dislocations, precipitates, and voids were characterized by TEM. Scanning transmission electron microscopy with energy dispersive X-ray spectroscopy (STEM-EDS) and/or atom probe tomography (APT) were used to study radiation-induced segregation and precipitation. Nanoindentation was used for hardness measurements to study irradiation hardening. Austenitic A709 and 316H was bulk-irradiated by 3.5 MeV Fe ++ ions to up to 150 peak dpa at 400, 500, and 600°. Compared to neutron-irradiated stainless steel (SS) 316, the Frank loop density of ion-irradiated A709 shows similar dose dependence at 400°, but very different temperature dependence. Due to the noticeable difference in the initial microstructure of A709 and 316H, no systematic comparison on the Frank loops in A709 vs 316H was made. It would be helpful that future ion irradiation study on 316 stainless steel could be conducted to directly compare the temperature dependence of Frank loop density in ion-irradiated 316 SS with that in neutron-irradiated 316 SS. In addition, future neutron irradiation on A709 at 400–600° at relative high dose (≥10 dpa) can be carried out to compare with ion-irradiated A709. The radiation

Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Long, Fei; Daymond, Mark R.; Yao, Zhongwen

2015-03-14

The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic < a > dislocations has been dynamically observed before and after irradiation at room temperature and 300 degrees C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by < a > dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting themore » possibility of basal channel formation in bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, {01 (1) over bar1}< 0 (1) over bar 12 > twinning was identified in the irradiated sample deformed at 300 degrees C.« less

Ion irradiation studies on the void swelling behavior of a titanium modified D9 alloy

NASA Astrophysics Data System (ADS)

Balaji, S.; Mohan, Sruthi; Amirthapandian, S.; Chinnathambi, S.; David, C.; Panigrahi, B. K.

2015-12-01

The sensitivity of Positron Annihilation Spectroscopy (PAS) for probing vacancy defects and their environment is well known. Its applicability in determination of swelling and the peak swelling temperature was put to test in our earlier work on ion irradiated D9 alloys [1]. Upon comparison with the peak swelling temperature determined by conventional step height measurements it was found that the peak swelling temperature determined using PAS was 50 K higher. It was conjectured that the positrons trapping in the irradiation induced TiC precipitation could have caused the shift. In the present work, D9 alloys have been implanted with 100 appm helium ions and subsequently implanted with 2.5 MeV Ni ions up to peak damage of 100 dpa. The nickel implantations have been carried out through a range of temperatures between 450 °C and 650 °C. The evolution of cavities and TiC precipitates at various temperatures has been followed by TEM and this report provides an experimental verification of the conjecture.

Irradiation-induced precipitation and mechanical properties of vanadium alloys at <430 C

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, H.M.; Gazda, J.; Smith, D.L.

Recent attention to V-base alloys has focused on the effect of low-temperature (<430 C) irradiation on tensile and impact properties of V-4Cr-4Ti. In previous studies, dislocation channeling, which causes flow localization and severe loss of work-hardening capability, has been attributed to dense, irradiation-induced precipitation of very fine particles. However, efforts to identify the precipitates were unsuccessful until now. In this study, analysis by transmission electron microscopy (TEM) was conducted on unalloyed V, V-5Ti, V-3Ti-1Si, and V-4Cr-4Ti specimens that were irradiated at <430 C in conventional and dynamic helium charging experiments. By means of dark-field imaging and selected-area-diffraction analysis, the characteristicmore » precipitates were identified to be (V,Ti{sub 1{minus}x})(C,O,N). In V-3Ti-1Si, precipitation of (V,Ti{sub 1{minus}x})(C,O,N) was negligible at <430 C, and as a result, dislocation channeling did not occur and work-hardening capability was high.« less

Proton irradiation damage of an annealed Alloy 718 beam window

DOE PAGES

Bach, H. T.; Anderoglu, O.; Saleh, T. A.; ...

2015-04-01

Mechanical testing and microstructural analysis was performed on an Alloy 718 window that was in use at the Los Alamos Neutron Science Center (LANSCE) Isotope Production Facility (IPF) for approximately 5 years. It was replaced as part of the IPF preventive maintenance program. The window was transported to the Wing 9 hot cells at the Chemical and Metallurgical Research (CMR) LANL facility, visually inspected and 3-mm diameter samples were trepanned from the window for mechanical testing and microstructural analysis. Shear punch testing and optical metallography was performed at the CMR hot cells. The 1-mm diameter shear punch disks were cutmore » into smaller samples to further reduce radiation exposure dose rate using Focus Ion Beam (FIB) and microstructure changes were analyzed using a Transmission Electron Microscopy (TEM). Irradiation doses were determined to be ~0.2–0.7 dpa (edge) to 11.3 dpa (peak of beam intensity) using autoradiography and MCNPX calculations. The corresponding irradiation temperatures were calculated to be ~34–120 °C with short excursion to be ~47–220 °C using ANSYS. Mechanical properties and microstructure analysis results with respect to calculated dpa and temperatures show that significant work hardening occurs but useful ductility still remains. The hardening in the lowest dose region (~0.2–0.7 dpa) was the highest and attributed to the formation of γ" precipitates and irradiation defect clusters/bubbles whereas the hardening in the highest dose region (~11.3 dpa) was lower and attributed mainly to irradiation defect clusters and some thermal annealing.« less

PROCESS FOR CONTINUOUSLY SEPARATING IRRADIATION PRODUCTS OF THORIUM

DOEpatents

Hatch, L.P.; Miles, F.T.; Sheehan, T.V.; Wiswall, R.H.; Heus, R.J.

1959-07-01

A method is presented for separating uranium-233 and protactinium from thorium-232 containing compositions which comprises irradiating finely divided particles of said thorium with a neutron flux to form uranium-233 and protactinium, heating the neutron-irradiated composition in a fluorine and hydrogen atmosphere to form volatile fluorides of uranium and protactinium and thereafter separating said volatile fluorides from the thorium.

Corrosion-resistant uranium

DOEpatents

Hovis, V.M. Jr.; Pullen, W.C.; Kollie, T.G.; Bell, R.T.

1981-10-21

The present invention is directed to the protecting of uranium and uranium alloy articles from corrosion by providing the surfaces of the articles with a layer of an ion-plated metal selected from aluminum and zinc to a thickness of at least 60 microinches and then converting at least the outer surface of the ion-plated layer of aluminum or zinc to aluminum chromate or zinc chromate. This conversion of the aluminum or zinc to the chromate form considerably enhances the corrosion resistance of the ion plating so as to effectively protect the coated article from corrosion.

Corrosion-resistant uranium

DOEpatents

Hovis, Jr., Victor M.; Pullen, William C.; Kollie, Thomas G.; Bell, Richard T.

1983-01-01

The present invention is directed to the protecting of uranium and uranium alloy articles from corrosion by providing the surfaces of the articles with a layer of an ion-plated metal selected from aluminum and zinc to a thickness of at least 60 microinches and then converting at least the outer surface of the ion-plated layer of aluminum or zinc to aluminum chromate or zinc chromate. This conversion of the aluminum or zinc to the chromate form considerably enhances the corrosion resistance of the ion plating so as to effectively protect the coated article from corrosion.

Microstructure and Elevated Temperature Properties of a Refractory TaNbHfZrTi Alloy

DTIC Science & Technology

2012-01-24

composition of the TaNbHfZrTi alloy produced by vacuum arc melting Composition Ta Nb Hf Zr Ti at.% 19.68 18.93 20.46 21.23 19.7 wt. % 30.04 14.84 30.82 16.34...metallic materials with higher melting points, such as refractory molybdenum (Mo) and niobium ( Nb ) alloys, are examined as alternatives by academic and...creep resistance are the key properties of these alloys, since considerable alloy softening generally occurs at tempera- tures above *0.5 0.6 Tm

Method for improving the mechanical properties of uranium-1 to 3 wt % zirconium alloy

DOEpatents

Anderson, R.C.

1983-11-22

A uranium-1 to 3 wt % zirconium alloy characterized by high strength, high ductility and stable microstructure is fabricated by an improved thermal mechanical process. A homogenous ingot of the alloy which has been reduced in thickness of at least 50% in the two-step forging operation, rolled into a plate with a 75% reduction and then heated in vacuum at a temperature of about 750 to 850/sup 0/C and then quenched in water, is subjected to further thermal-mechanical operation steps to increase the compressive yield strength approximately 30%, stabilize the microstructure, and decrease the variations in mechanical properties throughout the plate is provided. These thermal-mechanical steps are achieved by cold rolling the quenchd plate to reduce the thickness thereof about 8 to 12%, aging the cold rolled plate at a first temperature of about 325 to 375/sup 0/C for five to six hours and then aging the plate at a higher temperature ranging from 480 to 500/sup 0/C for five to six hours prior to cooling the billet to ambient conditions and sizing the billet or plate into articles provides the desired increase in mechanical properties and phase stability throughout the plate.

ALLOY COATINGS AND METHOD OF APPLYING

DOEpatents

Eubank, L.D.; Boller, E.R.

1958-08-26

A method for providing uranium articles with a pro tective coating by a single dip coating process is presented. The uranium article is dipped into a molten zinc bath containing a small percentage of aluminum. The resultant product is a uranium article covered with a thin undercoat consisting of a uranium-aluminum alloy with a small amount of zinc, and an outer layer consisting of zinc and aluminum. The article may be used as is, or aluminum sheathing may then be bonded to the aluminum zinc outer layer.

Gel Fabrication of Molybdenum “Beads”

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lowden, Richard Andrew; Armstrong, Beth L.; Cooley, Kevin M.

2016-11-01

Spherical molybdenum particles or “beads” of various diameters are of interest as feedstock materials for the additive manufacture of targets and assemblies used in the production of 99Mo medical isotopes using accelerator technology. Small metallic beads or ball bearings are typically fabricated from wire; however, small molybdenum spheres cannot readily be produced in this manner. Sol-gel processes are often employed to produce small dense microspheres of metal oxides across a broad diameter range that in the case of molybdenum could be reduced and sintered to produce metallic spheres. These Sol-gel type processes were examined for forming molybdenum oxide beads; however,more » the molybdenum trioxide was chemically incompatible with commonly used gelation materials. As an alternative, an aqueous alginate process being assessed for the fabrication of oxide spheres for catalyst applications was employed to form molybdenum trioxide beads that were successfully reduced and sintered to produce small molybdenum spheres.« less

Evaluation of a measurement system for Uranium electrodeposition control to radiopharmaceuticals production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tufic Madi Filho; Adonis Marcelo Saliba Silva; Jose Patricio Nahuel Cardenas

2015-07-01

For 2016, studies by international bodies forecast a crisis in the supply of Molybdenum ({sup 99}Mo), which is the generator of {sup 99m}Tc, widely used for medical diagnoses and treatments. As a result, many countries are making efforts to prevent this crisis. Brazil is developing the Brazilian Multipurpose Reactor (RMB) project, under the responsibility of the National Nuclear Energy Commission (CNEN). The RMB is a nuclear reactor for research and production of radioisotopes used in the production of radiopharmaceuticals and radioactive sources, broadly used in industrial and research areas in Brazil. Electrodeposition of uranium is a common practice to createmore » samples for alpha spectrometry and this methodology may be an alternative way to produce targets of low enriched uranium (LEU) to fabricate radiopharmaceuticals, as {sup 99}Mo, used for cancer diagnosis. To study the electrodeposition, a solution of 10 mM uranyl nitrate, in 2-propanol, containing uranium enriched to 2.4% in {sup 235}U, with pH = 1, was prepared and measurements with an alpha spectrometer were performed. These studies are justified by the need to produce {sup 99}Mo since, despite using molybdenum in bulk, Brazil is totally dependent on its import. In this project, we intend to obtain a process that may be technologically feasible to control the radiation targets for {sup 99}Mo production. (authors)« less

Effects of temperature on the irradiation responses of Al 0.1 CoCrFeNi high entropy alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, Tengfei; Xia, Songqin; Guo, Wei

Structural damage and chemical segregation in Al 0.1CoCrFeNi high entropy alloy irradiated at elevated temperatures are studied using transmission electron microscopy (TEM) and atom probe tomography (APT). Irradiation-induced defects include dislocation loops, long dislocations and stacking-fault tetrahedra, but no voids can be observed. Furthermore, as irradiation temperature increases, defect density is decreased but defect size is increased, which is induced by increasing defect mobility. Finally, APT characterization reveals that ion irradiation at elevated temperatures can induce an enrichment of Ni and Co as well as a depletion of Fe and Cr at defect clusters, mainly including dislocation loops and longmore » dislocations.« less

Effects of temperature on the irradiation responses of Al 0.1 CoCrFeNi high entropy alloy

DOE PAGES

Yang, Tengfei; Xia, Songqin; Guo, Wei; ...

2017-09-29

Structural damage and chemical segregation in Al 0.1CoCrFeNi high entropy alloy irradiated at elevated temperatures are studied using transmission electron microscopy (TEM) and atom probe tomography (APT). Irradiation-induced defects include dislocation loops, long dislocations and stacking-fault tetrahedra, but no voids can be observed. Furthermore, as irradiation temperature increases, defect density is decreased but defect size is increased, which is induced by increasing defect mobility. Finally, APT characterization reveals that ion irradiation at elevated temperatures can induce an enrichment of Ni and Co as well as a depletion of Fe and Cr at defect clusters, mainly including dislocation loops and longmore » dislocations.« less

Molybdenum reduction to molybdenum blue in Serratia sp. Strain DRY5 is catalyzed by a novel molybdenum-reducing enzyme.

PubMed

Shukor, M Y; Halmi, M I E; Rahman, M F A; Shamaan, N A; Syed, M A

2014-01-01

The first purification of the Mo-reducing enzyme from Serratia sp. strain DRY5 that is responsible for molybdenum reduction to molybdenum blue in the bacterium is reported. The monomeric enzyme has an apparent molecular weight of 105 kDalton. The isoelectric point of this enzyme was 7.55. The enzyme has an optimum pH of 6.0 and maximum activity between 25 and 35°C. The Mo-reducing enzyme was extremely sensitive to temperatures above 50°C (between 54 and 70°C). A plot of initial rates against substrate concentrations at 15 mM 12-MP registered a V max for NADH at 12.0 nmole Mo blue/min/mg protein. The apparent K m for NADH was 0.79 mM. At 5 mM NADH, the apparent V max and apparent K m values for 12-MP of 12.05 nmole/min/mg protein and 3.87 mM, respectively, were obtained. The catalytic efficiency (k cat/K m ) of the Mo-reducing enzyme was 5.47 M(-1) s(-1). The purification of this enzyme could probably help to solve the phenomenon of molybdenum reduction to molybdenum blue first reported in 1896 and would be useful for the understanding of the underlying mechanism in molybdenum bioremediation involving bioreduction.

Molybdenum Reduction to Molybdenum Blue in Serratia sp. Strain DRY5 Is Catalyzed by a Novel Molybdenum-Reducing Enzyme

PubMed Central

Shukor, M. Y.; Halmi, M. I. E.; Rahman, M. F. A.; Shamaan, N. A.; Syed, M. A.

2014-01-01

The first purification of the Mo-reducing enzyme from Serratia sp. strain DRY5 that is responsible for molybdenum reduction to molybdenum blue in the bacterium is reported. The monomeric enzyme has an apparent molecular weight of 105 kDalton. The isoelectric point of this enzyme was 7.55. The enzyme has an optimum pH of 6.0 and maximum activity between 25 and 35°C. The Mo-reducing enzyme was extremely sensitive to temperatures above 50°C (between 54 and 70°C). A plot of initial rates against substrate concentrations at 15 mM 12-MP registered a V max for NADH at 12.0 nmole Mo blue/min/mg protein. The apparent K m for NADH was 0.79 mM. At 5 mM NADH, the apparent V max and apparent K m values for 12-MP of 12.05 nmole/min/mg protein and 3.87 mM, respectively, were obtained. The catalytic efficiency (k cat/K m) of the Mo-reducing enzyme was 5.47 M−1 s−1. The purification of this enzyme could probably help to solve the phenomenon of molybdenum reduction to molybdenum blue first reported in 1896 and would be useful for the understanding of the underlying mechanism in molybdenum bioremediation involving bioreduction. PMID:24724104

Ultrasonic attenuation in superconducting molybdenum-rhenium alloys.

NASA Technical Reports Server (NTRS)

Ashkin, M.; Deis, D. W.; Gottlieb, M.; Jones, C. K.

1971-01-01

Investigation of longitudinal sound attenuation in superconducting Mo-Re alloys as a function of temperature, magnetic field, and frequency. Evaporated thin film CdS transducers were used for the measurements at frequencies up to 3 GHz. The normal state attenuation coefficient was found to be proportional to the square of frequency over this frequency range. Measurements in zero magnetic field yielded a value of the energy gap parameter close to the threshold value of 3.56 kTc, appropriate to a weakly coupled dirty limit superconductor.

Distribution of trace elements in drilling chip samples around a roll-type uranium deposit, San Juan Basin, New Mexico

USGS Publications Warehouse

Day, H.C.; Spirakis, C.S.; Zech, R.S.; Kirk, A.R.

1983-01-01

Chip samples from rotary drilling in the vicinity of a roll-type uranium deposit in the southwestern San Juan Basin were split into a whole-washed fraction, a clay fraction, and a heavy mineral concentrate fraction. Analyses of these fractions determined that cutting samples could be used to identify geochemical halos associated with this ore deposit. In addition to showing a distribution of selenium, uranium, vanadium, and molybdenum similar to that described by Harshman (1974) in uranium roll-type deposits in Wyoming, South Dakota, and Texas, the chemical data indicate a previously unrecognized zinc anomaly in the clay fraction downdip of the uranium ore.

Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Long, Fei; Daymond, Mark R., E-mail: mark.daymond@queensu.ca; Yao, Zhongwen

2015-03-14

The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic 〈a〉 dislocations has been dynamically observed before and after irradiation at room temperature and 300 °C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by 〈a〉 dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting the possibility of basal channel formation inmore » bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, (011{sup ¯}1)〈01{sup ¯}12〉 twinning was identified in the irradiated sample deformed at 300 °C.« less

Progress on the chemical separation of fission fragments from 236Np produced by proton irradiation of natural uranium target

NASA Astrophysics Data System (ADS)

Larijani, C.; Jerome, S. M.; Lorusso, G.; Ivanov, P.; Russell, B.; Pearce, A. K.; Regan, P. H.

2017-11-01

The aim of the current work is to develop and validate a radiochemical separation scheme capable of separating both 236gNp and 236Pu from a uranium target of natural isotopic composition ( 1 g uranium) and 200 MBq of fission decay products. A target containing 1.2 g of UO2 was irradiated with a beam of 25 MeV protons with a typical beam current of 30 μA for 19 h in December 2013 at the University of Birmingham Cyclotron facility. Using literature values for the production cross-section for fusion of protons with uranium targets, we estimate that an upper limit of approximately 250 Bq of activity from the 236Np ground state was produced in this experiment. Using a radiochemical separation scheme, Np and Pu fractions were separated from the produced fission decay products, with analyses of the target-based final reaction products made using Inductively Couple Plasma Mass Spectrometry (ICP-MS) and high-resolution α particle and γ-ray spectrometry.

Genetics Home Reference: molybdenum cofactor deficiency

MedlinePlus

... called molybdenum cofactor. Molybdenum cofactor, which contains the element molybdenum, is essential to the function of several ... Citation on PubMed or Free article on PubMed Central Reiss J, Gross-Hardt S, Christensen E, Schmidt P, ...

Method of producing molybdenum-99

DOEpatents

Pitcher, Eric John

2013-05-28

Method of producing molybdenum-99, comprising accelerating ions by means of an accelerator; directing the ions onto a metal target so as to generate neutrons having an energy of greater than 10 MeV; directing the neutrons through a converter material comprising techentium-99 to produce a mixture comprising molybdenum-99; and, chemically extracting the molybdenum-99 from the mixture.

Effect of solute atoms on swelling in Ni alloys and pure Ni under He + ion irradiation

NASA Astrophysics Data System (ADS)

Wakai, E.; Ezawa, T.; Imamura, J.; Takenaka, T.; Tanabe, T.; Oshima, R.

2002-12-01

The effects of solute atoms on microstructural evolutions have been investigated using Ni alloys under 25 keV He + irradiation at 500 °C. The specimens used were pure Ni, Ni-Si, Ni-Co, Ni-Cu, Ni-Mn and Ni-Pd alloys with different volume size factors. The high number densities of dislocation loops about 1.5×10 22 m -3 were formed in the specimens irradiated to 1×10 19 ions/m 2, and they were approximately equivalent, except for Ni-Si. The mean size of loops tended to increase with the volume size factor of solute atoms. In a dose of 4×10 20 ions/m 2, the swelling was changed from 0.2% to 4.5%, depending on the volume size factors. The number densities of bubbles tended to increase with the absolute values of the volume size factor, and the swelling increased with the volume size factors. This suggests that the mobility of helium and vacancy atoms may be influenced by the interaction of solute atoms with them.

PLURAL METALLIC COATINGS ON URANIUM AND METHOD OF APPLYING SAME

DOEpatents

Gray, A.G.

1958-09-16

A method is described of applying protective coatings to uranlum articles. It consists in applying chromium plating to such uranium articles by electrolysis in a chromic acid bath and subsequently applying, to this minum containing alloy. This aluminum contalning alloy (for example one of aluminum and silicon) may then be used as a bonding alloy between the chromized surface and an aluminum can.

Effects of as-cast and wrought Cobalt-Chrome-Molybdenum and Titanium-Aluminium-Vanadium alloys on cytokine gene expression and protein secretion in J774A.1 macrophages.

PubMed

Jakobsen, Stig S; Larsen, A; Stoltenberg, M; Bruun, J M; Soballe, K

2007-09-11

Insertion of metal implants is associated with a possible change in the delicate balance between pro- and anti-inflammatory proteins, probably leading to an unfavourable predominantly pro-inflammatory milieu. The most likely cause is an inappropriate activation of macrophages in close relation to the metal implant and wear-products. The aim of the present study was to compare surfaces of as-cast and wrought Cobalt-Chrome-Molybdenum (CoCrMo) alloys and Titanium-Aluminium-Vanadium (TiAlV) alloy when incubated with mouse macrophage J774A.1 cell cultures. Changes in pro- and anti-inflammatory cytokines (TNF-alpha, IL-6, IL-alpha, IL-1beta, IL-10) and proteins known to induce proliferation (M-CSF), chemotaxis (MCP-1) and osteogenesis (TGF-beta, OPG) were determined by ELISA and Real Time reverse transcriptase - PCR (Real Time rt-PCR). Lactate dehydrogenase (LDH) was measured in the medium to asses the cell viability. Surface properties of the discs were characterised with a profilometer and with energy dispersive X-ray spectroscopy. We here report, for the first time, that the prosthetic material surface (non-phagocytable) of as-cast high carbon CoCrMo reduces the pro-inflammatory cytokine IL-6 transcription, the chemokine MCP-1 secretion, and M-CSF secretion by 77%, 36%, and 62%, respectively. Furthermore, we found that reducing surface roughness did not affect this reduction. The results suggest that as-cast CoCrMo alloy is more inert than wrought CoCrMo and wrought TiAlV alloys and could prove to be a superior implant material generating less inflammation which might result in less osteolysis.

Rolling Process Modeling Report. Finite-Element Model Validation and Parametric Study on various Rolling Process parameters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soulami, Ayoub; Lavender, Curt A.; Paxton, Dean M.

2015-06-15

Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum alloy plate-type fuel for high-performance research reactors in the United States. This work supports the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization Reactor Conversion Program. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll-separating forces for various rolling mill geometries for PNNL, Babcock & Wilcox Co., Y-12 National Security Complex, Los Alamos National Laboratory, and Idaho National Laboratory. The model developed and presented in a previous report has been subjected to further validationmore » study using new sets of experimental data generated from a rolling mill at PNNL. Simulation results of both hot rolling and cold rolling of uranium-10% molybdenum coupons have been compared with experimental results. The model was used to predict roll-separating forces at different temperatures and reductions for five rolling mills within the National Nuclear Security Administration Fuel Fabrication Capability project. This report also presents initial results of a finite-element model microstructure-based approach to study the surface roughness at the interface between zirconium and uranium-10% molybdenum.« less

National Uranium Resource Evaluation Program. Hydrogeochemical and stream sediment reconnaissance basic data for Beeville NTMS Quadrangle, Texas. Uranium resource evaluation project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

Results of a reconnaissance geochemical survey of the Beeville Quadrangle, Texas are reported. Field and laboratory data are presented for 373 groundwater and 364 stream sediment samples. Statistical and areal distributions of uranium and possible uranium-related variables are displayed. A generalized geologic map of the survey area is provided, and pertinent geologic factors which may be of significance in evaluating the potential for uranium mineralization are briefly discussed. The groundwater data indicate that the northwestern corner of the quadrangle is the most favorable for potential uranium mineralization. Favorability is indicated by high uranium concentrations; high arsenic, molybdenum, and vanadium concentrations;more » and proximity and similar geologic setting to the mines of the Karnes County mining district. Other areas that appear favorable are an area in Bee and Refugio Counties and the northeastern part of the quadrangle. Both areas have water chemistry similar to the Karnes County area, but the northeastern area does not have high concentrations of pathfinder elements. The stream sediment data indicate that the northeastern corner of the quadrangle is the most favorable for potential mineralization, but agricultural practices and mineralogy of the outcropping Beaumont Formation may indicate a false anomaly. The northwestern corner of the quadrangle is considered favorable because of its proximity to the known uranium deposits, but the data do not seem to support this.« less

Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys.

PubMed

Lu, Chenyang; Jin, Ke; Béland, Laurent K; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M; Stoller, Roger E; Wang, Lumin

2016-02-01

Energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters far exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance.

Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys

PubMed Central

Lu, Chenyang; Jin, Ke; Béland, Laurent K.; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M.; Stoller, Roger E.; Wang, Lumin

2016-01-01

Energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters far exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance. PMID:26829570