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.

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.

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

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.

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

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.

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.

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.

Alloy hardening and softening in binary molybdenum alloys as related to electron concentration

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 hafnium, tantalum, tungsten, rhenium, osmium, iridium, and platinum on hardness of molybdenum. Special emphasis was placed on alloy softening in these binary molybdenum alloys. Results showed that alloy softening was produced by those elements having an excess of s+d electrons compared to molybdenum, while those elements having an equal number or fewer s+d electrons that molybdenum failed to produce alloy softening. Alloy softening and alloy hardening can be correlated with the difference in number of s+d electrons of the solute element and molybdenum.

Molybdenum-UO2 cermet irradiation at 1145 K.

NASA Technical Reports Server (NTRS)

Mcdonald, G.

1971-01-01

Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.

FY16 Status Report for the Uranium-Molybdenum Fuel Concept

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

Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.

2016-09-22

The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal yearmore » 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.« less

Mechanical properties of electron-beam-melted molybdenum and dilute molybdenum-rhenium alloys

NASA Technical Reports Server (NTRS)

Klopp, W. D.; Witzke, W. R.

1972-01-01

A study of molybdenum and three dilute molybdenum-rhenium alloys was undertaken to determine the effects of rhenium on the low temperature ductility and other mechanical properties of molybdenum. Alloys containing 3.9, 5.9, and 7.7 atomic percent rhenium exhibited lower ductile-brittle transition temperatures than did the unalloyed molybdenum. The maximum improvement in the annealed condition was observed for molybdenum - 7.7 rhenium, which had a ductile-brittle transition temperature approximately 200 C (360 F) lower than that for unalloyed molybdenum. Rhenium additions also increased the low and high temperature tensile strengths and the high temperature creep strength of molybdenum. The mechanical behavior of dilute molybdenum-rhenium alloys is similar to that observed for dilute tungsten-rhenium alloys.

Oxide strengthened molybdenum-rhenium alloy

DOEpatents

Bianco, Robert; Buckman, Jr., R. William

2000-01-01

Provided is a method of making an ODS molybdenum-rhenium alloy which includes the steps of: (a) forming a slurry containing molybdenum oxide and a metal salt dispersed in an aqueous medium, the metal salt being selected from nitrates or acetates of lanthanum, cerium or thorium; (b) heating the slurry in the presence of hydrogen to form a molybdenum powder comprising molybdenum and an oxide of the metal salt; (c) mixing rhenium powder with the molybdenum powder to form a molybdenum-rhenium powder; (d) pressing the molybdenum-rhenium powder to form a molybdenum-rhenium compact; (e) sintering the molybdenum-rhenium compact in hydrogen or under a vacuum to form a molybdenum-rhenium ingot; and (f) compacting the molybdenum-rhenium ingot to reduce the cross-sectional area of the molybdenum-rhenium ingot and form a molybdenum-rhenium alloy containing said metal oxide. The present invention also provides an ODS molybdenum-rhenium alloy made by the method. A preferred Mo--Re-ODS alloy contains 7-14 weight % rhenium and 2-4 volume % lanthanum oxide.

NPF MECHANICAL CELL NaK DISPOSAL AND FUME ABATEMENT

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

Rey, G.

Some of the fuels originally scheduled for processing in the nonproduction fuel (NPF) processing program incorporated sodium or sodium- potassium alloy (NaK) as the bonding material between stainless-steel cladding and the uranium or uranium-molybdenum alloy core. Because of the special hazards involved in handling NaK, studies were made to determine safe methods for processing NaK-containing fuels. An underwater NaK dispensing system was installed, and tests were made to determine the characteristics of the NaK-water reaction. The equipment consisted of a dispenser, reaction pan, and off-gas scrubber. After initinl studies, a prototype test was made wherein U-Mo canned slugs containing NaKmore » reservoirs were hack sawed underwater. The studies demonstrated that the NaK reservoirs can be safely deactivated by hack sawing under a submerged hood in a shallow water bath. (W.L.H.)« less

Recovering and recycling uranium used for production of molybdenum-99

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

Reilly, Sean Douglas; May, Iain; Copping, Roy

A processes for recycling uranium that has been used for the production of molybdenum-99 involves irradiating a solution of uranium suitable for forming fission products including molybdenum-99, conditioning the irradiated solution to one suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina. Another process involves irradiation of a solid target comprising uranium, forming an acidic solution from the irradiated targetmore » suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina.« less

Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target

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

Reilly, Sean Douglas; May, Iain; Copping, Roy

A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted tomore » concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.« less

Manufacturing Experience for Oxide Dispersion Strengthened Alloys

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

Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.

2016-09-22

This report documents the results of the development and the manufacturing experience gained at the Pacific Northwest National Laboratories (PNNL) while working with the oxide dispersion strengthened (ODS) materials MA 956, 14YWT, and 9YWT. The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. ODS materials have the potential to provide improved performance for the U-Mo concept.

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.

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.

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.

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, and intergranular corrosion does not take place. In the fuel salt with [U(IV)]/[U(III)] = 4-20 the potentials of uranium alloy formation with the main components of the tested alloys are not reached, that's why alloys and intermetallic compounds are not formed on the surface of the investigated chromium-nickel alloys. Under such conditions any intergranular tellurium corrosion of the selected alloys does not occur. In the fuel salt with [U(IV)/]/[U(III)] = 100 the potentials of uranium alloy formation with the main components of the tested alloys are not also reached. Under such redox conditions any traces intergranular tellurium IGC on the HN80MTY and H80M-VI alloys specimens are not found. Certain signs of incipient IGC in the form of tellurium presence on the grain boundaries in the HN80MTB and EM-721 alloys surface layer and formation of not too deep cracks on HN80MTB alloy surface were revealed at [U(IV)/]/[U(III)] = 100. With this uranium ratio in the presence of corrosion products on the surface of all of the alloys films, containing tellurium, metals of the construction alloys and carbon, are formed. In the melt with [U(IV)]/[U(III)] = 500 in all of the alloys tested the tellurium IGC took place. The HN80MTY alloy shows the maximum resistance to tellurium IGC. The intensity of tellurium IGC of the alloy (the K parameter) is by 3-5 times lower as compared to other alloys. The EM-721 alloy has the minimal resistance to tellurium IGC (K = 9200 pc m/cm, the depth of cracks is up to 434 μm). The studies have shown, that the intensity of the nickel alloys IGC is controlled by the [U(IV)]/[U(III)] ratio, and its dependence on this parameter is of threshold character. Providing the uranium ratio value's monitoring and regulation, it is possible to control the tellurium corrosion and in such a way to eliminate IGC completely or to minimize its value. The alloys strength characteristics and their structure were changed insignificantly after testing within the [U(IV)]/[U(III)] range from 0.7 tо 100. The changes are not linked with the influence of fuel salt, containing tellurium additions, but are stipulated by alloys structure, temperature factor, exposure time and mechanical loads. Significant effect of tellurium cracking on the alloys (excepting HN80MTY) strength characteristics was established after corrosion testing with [U(IV)]/[U(III)] = 500. In the absence of IGC all of the alloys investigated have a good ductility at high strength characteristics. The disrupture of specimens under mechanical tests both before and after corrosion tests of all alloys except for ЕМ-721 proceeds on a ductile mechanism. On the EM-721 alloy specimens, both in their initial state and after corrosion testing, clear signs of brittle destruction, caused by heterogeneity of its structure due to the presence of tungsten phase, are very clearly observed. The presence of such phases increases the alloy IGC and leads to reduction of the alloy resistance tellurium damage. The HN80MTY alloy has the best corrosion and mechanical properties. It does not undergo tellurium IGC in the molten 75LiF-5BeF2-20ThF4 salt mixture fueled by about 2 mol% of UF4 with [U(IV)]/[U(III)] ratio ⩽ 100. The alloy has high resistance to tellurium cracking at [U(IV)]/[U(III)] = 500. The alloy can be recommended as the main construction material for the fuel circuit with selected salt composition up to temperature 750 °С.

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.

Molybdenum-A Key Component of Metal Alloys

USGS Publications Warehouse

Kropschot, S.J.

2010-01-01

Molybdenum, whose chemical symbol is Mo, was first recognized as an element in 1778. Until that time, the mineral molybdenite-the most important source of molybdenum-was believed to be a lead mineral because of its metallic gray color, greasy feel, and softness. In the late 19th century, French metallurgists discovered that molybdenum, when alloyed (mixed) with steel in small quantities, creates a substance that is remarkably tougher than steel alone and is highly resistant to heat. The alloy was found to be ideal for making tools and armor plate. Today, the most common use of molybdenum is as an alloying agent in stainless steel, alloy steels, and superalloys to enhance hardness, strength, and resistance to corrosion.

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

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.

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.

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

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

Montgomery, Robert O.; Bennett, Wendy D.; Henager, Charles H.

2016-09-12

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impactmore » of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.« less

PURIFICATION PROCESS

DOEpatents

Wibbles, H.L.; Miller, E.I.

1958-01-14

This patent deals with the separation of uranium from molybdenum compounds, and in particular with their separation from ether solutions containing the molybdenum in the form of acids, such as silicomolybdic and phosphomolybdic acids. After the nitric acid leach of pitchblende, the molybdenum values present in the ore are found in the leach solution in the form of complex acids. The uranium bearing solution may be purified of this molybdenum content by comtacting it with activated charcoal. The purification is improved when the acidity of the solution is low ad agitation is also beneficial. The molybdenum may subsequently be recovered from the charcosl ad the charcoal reused.

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

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.

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

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.

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.

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

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

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.

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.

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

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

Diamond, D. J.; Baek, J. S.; Hanson, A. L.

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in anmore » aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.« less

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.

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.

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.

Marine Corrosion.

DTIC Science & Technology

1985-04-24

cor- rosion resistant alloys such as molybdenum -containing stainless steels. For the latter the high degree of aeration in the splashing water...imposed by marine technology, such as elevated temperatures , tensile stresses, cyclic stresses, severe (tight) crevices, galvanic coupling and high ...corrosion in seawater in tight metal-to-non-metal crevices are titanium alloys 4, the high molybdenum nickel base alloys Hastelloy alloy C-276 and

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.

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)

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.

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)

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.

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.

21 CFR 888.3390 - Hip joint femoral (hemi-hip) metal/polymer cemented or uncemented prosthesis.

Code of Federal Regulations, 2014 CFR

2014-04-01

... includes prostheses that have a femoral component made of alloys, such as cobalt-chromium-molybdenum, and a snap-fit acetabular component made of an alloy, such as cobalt-chromium-molybdenum, and ultra-high...

21 CFR 888.3390 - Hip joint femoral (hemi-hip) metal/polymer cemented or uncemented prosthesis.

Code of Federal Regulations, 2013 CFR

2013-04-01

... includes prostheses that have a femoral component made of alloys, such as cobalt-chromium-molybdenum, and a snap-fit acetabular component made of an alloy, such as cobalt-chromium-molybdenum, and ultra-high...

21 CFR 888.3390 - Hip joint femoral (hemi-hip) metal/polymer cemented or uncemented prosthesis.

Code of Federal Regulations, 2011 CFR

2011-04-01

... includes prostheses that have a femoral component made of alloys, such as cobalt-chromium-molybdenum, and a snap-fit acetabular component made of an alloy, such as cobalt-chromium-molybdenum, and ultra-high...

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.

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.

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.

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.

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

Stillman, J. A.; Feldman, E. E.; Wilson, E. H.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains themore » results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo (U-10Mo).« less

Hydrogeochemical and stream sediment detailed geochemical survey for Edgemont, South Dakota; Wyoming

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

Butz, T.R.; Dean, N.E.; Bard, C.S.

1980-05-31

Results of the Edgemont detailed geochemical survey are reported. Field and laboratory data are presented for 109 groundwater and 419 stream sediment samples. Statistical and areal distributions of uranium and possible uranium-related variables are given. 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. Groundwaters containing greater than or equal to 7.35 ppB uranium are present in scattered clusters throughout the area sampled. Most of these groundwaters are from wells drilled where the Inyan Kara Group is exposed at themore » surface. The exceptions are a group of samples in the northwestern part of the area sampled and south of the Dewey Terrace. These groundwaters are also produced from the Inyan Kara Group where it is overlain by the Graneros Group and alluvium. The high uranium groundwaters along and to the south of the terrace are characterized by high molybdenum, uranium/specific conductance, and uranium/sulfate values. Many of the groundwaters sampled along the outcrop of the Inyan Kara Group are near uranium mines. Groundwaters have high amounts of uranium and molybdenum. Samples taken downdip are sulfide waters with low values of uranium and high values of arsenic, molybdenum, selenium, and vanadium. Stream sediments containing greater than or equal to 5.50 ppM soluble uranium are concentrated in basins draining the Graneros and Inyan Kara Groups. These values are associated with high values for arsenic, selenium, and vanadium in samples from both groups. Anomalous values for these elements in the Graneros Group may be caused by bentonite beds contained in the rock units. As shown on the geochemical distribution plot, high uranium values that are located in the Inyan Kara Group are almost exclusively draining open-pit uranium mines.« less

Interdiffusion behaviors in doped molybdenum uranium and aluminum or aluminum silicon dispersion fuels: Effects of the microstructure

NASA Astrophysics Data System (ADS)

Allenou, J.; Tougait, O.; Pasturel, M.; Iltis, X.; Charollais, F.; Anselmet, M. C.; Lemoine, P.

2011-09-01

Si addition to Al is considered as a promising route to reduce (U,Mo)-Al interaction kinetics, due to its accumulation in the interaction layer, yielding the formation of silicide phases. The (U,Mo) alloy microstructure, and especially its homogenization state, could play a role on this accumulation process. The addition of a third element in γ(U,Mo) could also influence diffusion mechanisms of Al and Si. These two parameters were studied by means of diffusion couple experiments by joining γU based alloys with Al and (Al,Si) alloy. Chemical elements X added into γ(U,Mo) were thoroughly chosen on the following criteria: (i) the potential solubility of the alloying element into the γ(U,Mo) matrix, (ii) its capability to form the ternary aluminides based on the CeCr 2Al 20 and Ho 6Mo 4Al 43 - types, and (iii) the feasibility to control the microstructure of the alloys. On this basis, a test matrix is defined. It concerns γ(U80,Mo15,X5) alloys (in at.%) with X = Y, Cu, Zr, Ti or Cr. These alloys were homogenized and coupled with Al or (Al,Si) alloy. Results evidenced, first, the importance of the state of homogenization of the γ(U,Mo) binary alloy on interaction processes with (Al,Si) alloy, and the benefit on the diffusion of Si through the interaction layer, as observed on the elementary concentration profiles, when the third element X has some solubility into γ(U,Mo) alloy.

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.

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.

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.

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.

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.

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

Bruce S. Kang

The objective of this project was to understand and improve high-temperature structural properties of metal-silicide intermetallic alloys. Through research collaboration between the research team at West Virginia University (WVU) and Dr. J.H. Schneibel at Oak Ridge National Laboratory (ORNL), molybdenum silicide alloys were developed at ORNL and evaluated at WVU through atomistic modeling analyses, thermo-mechanical tests, and metallurgical studies. In this study, molybdenum-based alloys were ductilized by dispersing MgAl2O4 or MgO spinel particles. The addition of spinel particles is hypothesized to getter impurities such as oxygen and nitrogen from the alloy matrix with the result of ductility improvement. The introductionmore » of fine dispersions has also been postulated to improve ductility by acting as a dislocation source or reducing dislocation pile-ups at grain boundaries. The spinel particles, on the other hand, can also act as local notches or crack initiation sites, which is detrimental to the alloy mechanical properties. Optimization of material processing condition is important to develop the desirable molybdenum alloys with sufficient room-temperature ductility. Atomistic analyses were conducted to further understand the mechanism of ductility improvement of the molybdenum alloys and the results showed that trace amount of residual oxygen may be responsible for the brittle behavior of the as-cast Mo alloys. For the alloys studied, uniaxial tensile tests were conducted at different loading rates, and at room and elevated temperatures. Thermal cycling effect on the mechanical properties was also studied. Tensile tests for specimens subjected to either ten or twenty thermal cycles were conducted. For each test, a follow-up detailed fractography and microstructural analysis were carried out. The test results were correlated to the size, density, distribution of the spinel particles and processing time. Thermal expansion tests were carried out using thermo-mechanical analyzer (TMA). Results showed that the coefficient of thermal expansion (CTE) value decreases with the addition of spinel and silicide particles. Thermo-cycling tests showed that molybdenum alloy with 6% wt of spinel (MgAl2O4) developed microcracks which were caused by thermal expansion mismatch between the spinel particles and molybdenum matrix, as well as the processing conditions. Detailed post-mortem studies of microstructures and segregation of impurities to the oxide dispersion/Mo interfaces were conducted using x-ray diffraction (XRD), scanning electron microscopy (SEM), and energy dispersive spectroscopy (EDS).« less

Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts

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

Van Kleeck, M.; Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439; Willit, J.

A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling.more » In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl{sub 3}. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K{sub 2}UCl{sub 6} at the uranium surface, within the error of the analytical method.« 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.

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.

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

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.)

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

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.

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)

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).

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

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.

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.

EFFECTS OF COMPOSITION ON THE MECHANICAL PROPERTIES OF NI-CR-MO-CO FILLER METALS.

DTIC Science & Technology

STEEL, WELDING RODS), CHEMICAL ANALYSIS, CARBON ALLOYS , COBALT ALLOYS , CHROMIUM ALLOYS , MOLYBDENUM ALLOYS , NICKEL ALLOYS , MARAGING STEELS...ALUMINUM COMPOUNDS, TITANIUM , NONMETALS, SHIP HULLS, SHIP PLATES, SUBMARINE HULLS, WELDING , WELDS , MECHANICAL PROPERTIES, STATISTICAL ANALYSIS, MICROSTRUCTURE.

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.

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

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.

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.

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.

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.

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.

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

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

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

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.

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

High-Temperature Crystal-Growth Cartridge Tubes Made by VPS

NASA Technical Reports Server (NTRS)

Holmes, Richard; O'Dell, Scott; McKechnie, Timothy; Power, Christopher

2008-01-01

Cartridge tubes for use in a crystal growth furnace at temperatures as high as 1,600 deg. C have been fabricated by vacuum plasma spraying (VPS). These cartridges consist mainly of an alloy of 60 weight percent molybdenum with 40 weight percent rhenium, made from molybdenum powder coated with rhenium. This alloy was selected because of its high melting temperature (approximately equal.2,550 C) and because of its excellent ductility at room temperature. These cartridges are intended to supplant tungsten/nickel-alloy cartridges, which cannot be used at temperatures above approximately equal 1,300 C.

Thermal-desorption measurements for estimating bakeout characteristics of vacuum devices

NASA Astrophysics Data System (ADS)

Beavis, L.

1981-11-01

This discussion will be confined to outgassing phenomena; although gettering (sinks) or permeation (transfer through the entire vacuum wall) are imported in long term prediction. Measuring outgassing rates directly is complicated by the dynamic interaction between the samples being measured and the apparatus in which the measurements are made. Thermoesorption data are presented for molybdenum, nickel, Fe-Ni-Co alloy, copper, Cu-Be alloy, molybdenum sealing glass ceramic, and high-alumina ceramic.

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.

Study of the effects of gaseous environmental on the hot corrosion of superalloy materials

NASA Technical Reports Server (NTRS)

Smeggil, J. G.

1981-01-01

Studies have been conducted to examine the effect of low concentrations of NaCl(g) on the high temperature oxidation behavior of complex superalloys and potential coating formulations modified by silicon and reactive element (i.e., yttrium and hafnium) additions. Depending on alloy composition, a variety of effects were thermogravimetrically produced. Aluminum free alloys such as MAR-M509 and Hastelloy X with molybdenum and tungsten in solid solution showed accelerated (or breakaway) kinetics similar to that observed for Ni-Cr alloys. For IN-792, an alloy high in chromium and low in aluminum, molybdenum and tungsten present in solid solution does not adversely affect oxidation kinetics in the presence of NaCl(g). On the other hand, nickel-base alloys high in aluminum and molybdenum are catastrophically attacked by NaCl-bearing atmospheres. Silicon additions were, in general, observed to slightly improve the oxidation resistance of Ni, Ni-40Cr and CoCrAlY compositions in NaCl(g)-bearing atmospheres. To the degree that processes responsible for Al2O3 whisker formation deleteriously affect protective scale adherence, the addition of yttrium or hafnium can inhibit such whisker growth.

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

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.

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 recovery is examined, in part qualitatively and in part quantitatively, based upon the preceding data garnered through literature review, modeling, and experimentation. The sum of this research is meant to allow for a complete understanding of the process flow, from the beginning of one production cycle to the beginning of another.

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.

Development of explosively bonded TZM wire reinforced Columbian sheet composites

NASA Technical Reports Server (NTRS)

Otto, H. E.; Carpenter, S. H.

1972-01-01

Methods of producing TZM molybdenum wire reinforced C129Y columbium alloy composites by explosive welding were studied. Layers of TZM molybdenum wire were wound on frames with alternate layers of C129Y columbium alloy foil between the wire layers. The frames held both the wire and foils in place for the explosive bonding process. A goal of 33 volume percent molybdenum wire was achieved for some of the composites. Variables included wire diameter, foil thickness, wire separation, standoff distance between foils and types and amounts of explosive. The program was divided into two phases: (1) development of basic welding parameters using 5 x 10-inch composites, and (2) scaleup to 10 x 20-inch composites.

METAL COATED ARTICLES AND METHOD OF MAKING

DOEpatents

Eubank, L.D.

1958-08-26

A method for manufacturing a solid metallic uranium body having an integral multiple layer protective coating, comprising an inner uranium-aluminum alloy firmly bonded to the metallic uranium is presented. A third layer of silver-zinc alloy is bonded to the zinc-aluiminum layer and finally a fourth layer of lead-silver alloy is firmly bonded to the silver-zinc layer.

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.

Standard Method for Analyzing Gases in Titanium and Titanium Alloys. Standard Method for the Chemical Analysis of Titanium Alloys.

DTIC Science & Technology

1982-10-28

form a non- soluble complex. After filtering and burning the non-pure molybdenum trioxide is weighed. Ammonia water is used to dissolve the molybdenum...niobium and tantalum should use the methyl alcohol distillation - curcumin absorption luminosity 66 method for determination. II. The Methyl Alcohol...Distillation - Curcumin Absorption Luminosity Method 1. Summary of Method In a phosphorus sulfate medium, boron and methyl alcohol produce methyl borate

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.

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.

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.

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

Preliminary Material Properties Handbook, SI Units

DTIC Science & Technology

1999-12-01

5.5 Beta, Near-Beta, and Metastable Titanium Alloys 5-11 References 5-17 Chapter 6. Heat-Resistant Alloys 6.1 General 6-1 6.2 Iron- Chromium ...elements as vanadium, molybdenum, iron, or chromium . In addition to strengthening of titanium by the alloying additions, alpha-beta alloys may be...ALLOYS Heat-resistant alloys are arbitrarily defined as iron alloys richer in alloy content than the 18 percent chromium , 8 percent nickel types

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

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.

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 precipitation of uranium minerals. At the deepest exposed levels, wall-rocks were altered to sericite; and uraninite, coffinite, jordisite, fluorite, molybdenite, quartz, and pyrite were deposited in the veins. The fluids were progressively oxidized and cooled at higher levels in the system by boiling and degassing; iron-bearing minerals in wall rocks were oxidized to hematite, and quartz, fluorite, minor siderite, and uraninite were deposited in the veins. Near the ground surface, the fluids were acidified by condensation of volatiles and oxidation of hydrogen sulfide in near-surface, steam-heated, ground waters; wall rocks were altered to kaolinite, and quartz fluorite, and uraninite were deposited in veins. Secondary uranium minerals, hematite, and gypsum formed during supergene alteration later in the Cenozoic when the upper part of the mineralized system was exposed by erosion.

Physical, mechanical, and flexural properties of 3 orthodontic wires: an in-vitro study.

PubMed

Juvvadi, Shubhaker Rao; Kailasam, Vignesh; Padmanabhan, Sridevi; Chitharanjan, Arun B

2010-11-01

Understanding the biologic requirements of orthodontic patients requires proper characterization studies of new archwire alloys. The aims of this study were to evaluate properties of wires made of 2 new materials and to compare their properties with those of stainless steel. The sample consisted of 30 straight lengths of 3 types of wires: stainless steel, titanium-molybdenum alloy, and beta-titanium alloy. Eight properties were evaluated: wire dimension, edge bevel, composition, surface characteristics, frictional characteristics, ultimate tensile strength (UTS), modulus of elasticity (E), yield strength (YS), and load deflection characteristics. A toolmaker's microscope was used to measure the edge bevel, and x-ray fluorescence was used for composition analysis. Surface profilometry and scanning electron microscopy were used for surface evaluation. A universal testing machine was used to evaluate frictional characteristics, tensile strength, and 3-point bending. Stainless steel was the smoothest wire; it had the lowest friction and spring-back values and high values for stiffness, E, YS, and UTS. The titanium-molybdenum alloy was the roughest wire; it had high friction and intermediate spring-back, stiffness, and UTS values. The beta-titanium alloy was intermediate for smoothness, friction, and UTS but had the highest spring-back. The beta-titanium alloy with increased UTS and YS had a low E value, suggesting that it would have greater resistance to fracture, thereby overcoming a major disadvantage of titanium-molybdenum alloy wires. The beta-titanium alloy wire would also deliver gentler forces. Copyright © 2010 American Association of Orthodontists. Published by Mosby, Inc. All rights reserved.

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.

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.

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.

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 chemical weathering in terrestrial environments where the role of secondary processes such as adsorption is significant.

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

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.

Preliminary Material Properties Handbook, English Units

DTIC Science & Technology

1999-12-01

References 5-17 Chapter 6. Heat-Resistant Alloys 6.1 General 6-1 6.2 Iron- Chromium -Nickel-Base Alloys 6-3 6.3 Nickel-Base Alloys 6-3 6.4...elements as vanadium, molybdenum, iron, or chromium . In addition to strengthening of titanium by the alloying additions, alpha-beta alloys may be...alloys are arbitrarily defined as iron alloys richer in alloy content than the 18 percent chromium , 8 percent nickel types, or as alloys with a base

Method of making crack-free zirconium hydride

DOEpatents

Sullivan, Richard W.

1980-01-01

Crack-free hydrides of zirconium and zirconium-uranium alloys are produced by alloying the zirconium or zirconium-uranium alloy with beryllium, or nickel, or beryllium and scandium, or nickel and scandium, or beryllium and nickel, or beryllium, nickel and scandium and thereafter hydriding.

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.

76 FR 66684 - Foreign-Trade Zone 29-Louisville, KY, Application for Subzone; North American Stainless...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-27

...) include: fluorospar, molybdenum oxide, ferromanganese, ferrosilicon, ferrosilicon manganese, charge chrome... spent anodes, nickel, unwrought nickel alloys, aluminum, zinc, zinc alloys, manganese metal, titanium...

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.

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

NASA Astrophysics Data System (ADS)

Rest, J.; Hofman, G. L.; Kim, Yeon Soo

2009-04-01

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

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.

Geochemical reconnaissance study of Vassar Meadow (Adams Rib) wetlands and vicinity, Eagle County, Colorado

USGS Publications Warehouse

Owen, Douglass E.; Breit, George N.

1995-01-01

Wetlands are known to be efficient filters of metals dissolved in ground and surface waters. This paper presents the results of geochemical reconnaissance sampling done at the request of the U.S. Environmental Protection Agency in wetlands in Vassar Meadow, Eagle County, Colorado. Ten wetlands were sampled and found to be variously enriched in chromium, molybdenum, and uranium. The uranium and chromium concentrations (and, to a lesser extent, molybdenum) represent an environmental concern should they be released as a result of anthropogenic disturbance. The metal accumulation in these wetlands documents that the wetlands have been functioning as filters that protect water quality in East Brush Creek by lowering the dissolved metal content in water.

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.

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.

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.

Method for producing evaporation inhibiting coating for protection of silicon--germanium and silicon--molybdenum alloys at high temperatures in vacuum

DOEpatents

Chao, P.J.

1974-01-01

A method is given for protecting Si--Ge and Si-- Mo alloys for use in thermocouples. The alloys are coated with silicon to inhibit the evaporation of the alloys at high tempenatures in a vacuum. Specific means and methods are provided. (5 fig) (Official Gazette)

METAL COMPOSITIONS

DOEpatents

Seybolt, A.U.

1959-02-01

Alloys of uranium which are strong, hard, and machinable are presented, These alloys of uranium contain bctween 0.1 to 5.0% by weight of at least one noble metal such as rhodium, palladium, and gold. The alloys may be heat treated to obtain a product with iniproved tensile and compression strengths,

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

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.

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.

48 CFR 252.225-7008 - Restriction on Acquisition of Specialty Metals.

Code of Federal Regulations, 2012 CFR

2012-10-01

... atomization or sputtering of titanium, or final consolidation of non-melt derived titanium powder or titanium alloy powder. (3) Specialty metal means— (i) Steel— (A) With a maximum alloy content exceeding one or..., molybdenum, nickel, niobium (columbium), titanium, tungsten, or vanadium; (ii) Metal alloys consisting of— (A...

48 CFR 252.225-7008 - Restriction on Acquisition of Specialty Metals.

Code of Federal Regulations, 2010 CFR

2010-10-01

... atomization or sputtering of titanium, or final consolidation of non-melt derived titanium powder or titanium alloy powder. (3) Specialty metal means— (i) Steel— (A) With a maximum alloy content exceeding one or..., molybdenum, nickel, niobium (columbium), titanium, tungsten, or vanadium; (ii) Metal alloys consisting of— (A...

48 CFR 252.225-7008 - Restriction on Acquisition of Specialty Metals.

Code of Federal Regulations, 2011 CFR

2011-10-01

... atomization or sputtering of titanium, or final consolidation of non-melt derived titanium powder or titanium alloy powder. (3) Specialty metal means— (i) Steel— (A) With a maximum alloy content exceeding one or..., molybdenum, nickel, niobium (columbium), titanium, tungsten, or vanadium; (ii) Metal alloys consisting of— (A...

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.

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.

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

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.

Metals in Urine and Diabetes in U.S. Adults

PubMed Central

Guallar, Eliseo; Cowie, Catherine C.

2016-01-01

Our objective was to evaluate the relationship of urine metals including barium, cadmium, cobalt, cesium, molybdenum, lead, antimony, thallium, tungsten, and uranium with diabetes prevalence. Data were from a cross-sectional study of 9,447 participants of the 1999–2010 National Health and Nutrition Examination Survey, a representative sample of the U.S. civilian noninstitutionalized population. Metals were measured in a spot urine sample, and diabetes status was determined based on a previous diagnosis or an A1C ≥6.5% (48 mmol/mol). After multivariable adjustment, the odds ratios of diabetes associated with the highest quartile of metal, compared with the lowest quartile, were 0.86 (95% CI 0.66–1.12) for barium (Ptrend = 0.13), 0.74 (0.51–1.09) for cadmium (Ptrend = 0.35), 1.21 (0.85–1.72) for cobalt (Ptrend = 0.59), 1.31 (0.90–1.91) for cesium (Ptrend = 0.29), 1.76 (1.24–2.50) for molybdenum (Ptrend = 0.01), 0.79 (0.56–1.13) for lead (Ptrend = 0.10), 1.72 (1.27–2.33) for antimony (Ptrend < 0.01), 0.76 (0.51–1.13) for thallium (Ptrend = 0.13), 2.18 (1.51–3.15) for tungsten (Ptrend < 0.01), and 1.46 (1.09–1.96) for uranium (Ptrend = 0.02). Higher quartiles of barium, molybdenum, and antimony were associated with greater HOMA of insulin resistance after adjustment. Molybdenum, antimony, tungsten, and uranium were positively associated with diabetes, even at the relatively low levels seen in the U.S. population. Prospective studies should further evaluate metals as risk factors for diabetes. PMID:26542316

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.

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 physical description of fission product behavior fuels for advanced power reactors.

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

Kaganas, G.; Rest, J.; Nuclear Engineering Division

2007-10-18

The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuelsmore » under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.« less

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

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

Burkes, Douglas E.; Senor, David J.; Casella, Andrew M.

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 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. The current paper extends a failure model originally developed for UO2-stainless steel dispersion fuels and used currently available thermal-mechanical property information for the materials ofmore » interest in the current proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the 235U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the yield strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of interaction layer formation and can extend the performance of a fuel plate under a certain set of irradiation conditions, primarily moderate heat flux and burnup. Increasing the dispersed fuel particle diameter may also be effective, but only when combined with other parameters, e.g., lower enrichment and increased Si concentration. The model may serve as a valuable tool in initial experimental design.« less

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.

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.

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.

Histopathology of mallards dosed with lead and selected substitute shot

USGS Publications Warehouse

Locke, L.N.; Irby, H.D.; Bagley, George E.

1967-01-01

The histopathological response of male game farm mallards fed lead, three types of plastic-coated lead, two lead-magnesium alloys, iron, copper, zinc-coated iron, and molybdenum-coated iron shot was studied. Mallards fed lead, plastic-coated lead, or lead-magnesium alloy shot developed a similar pathological response, including the formation of acid-fast intranuclear inclusion bodies in the kidneys. Birds fed iron or molybdenum-coated iron shot developed hemosiderosis of the liver. Two of four mallards fed zinc-coated iron shot also developed hemosiderosis of the liver. No lesions were found in mallards fed copper shot.

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.

Electrocatalyst compositions

DOEpatents

Mallouk, Thomas E.; Chan, Benny C.; Reddington, Erik; Sapienza, Anthony; Chen, Guoying; Smotkin, Eugene; Gurau, Bogdan; Viswanathan, Rameshkrishnan; Liu, Renxuan

2001-09-04

Compositions for use as catalysts in electrochemical reactions are described. The compositions are alloys prepared from two or more elemental metals selected from platinum, molybdenum, osmium, ruthenium, rhodium, and iridium. Also described are electrode compositions including such alloys and electrochemical reaction devices including such catalysts.

Preliminary Material Properties Handbook. Volume 1: English Units

DTIC Science & Technology

2000-07-01

6-1 6.2 Iron- Chromium -Nickel-Base Alloys...titanium but is stabilized to room temperature by sufficient quantities of beta stabilizing elements as vanadium, molybdenum, iron, or chromium . In...Designation 6.2 6.3 6.3.1 6.3.2 6.3.3 6.3.4 6.3.5 6.4 6.5 6.5.1 Iron- Chromium -Nickel-Base Alloys Nickel-Base Alloys AEREX® 350 alloy HAYNES® 230® alloy

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

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.

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.

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.

Carbon-free induction furnace

DOEpatents

Holcombe, Cressie E.; Masters, David R.; Pfeiler, William A.

1985-01-01

An induction furnace for melting and casting highly pure metals and alloys such as uranium and uranium alloys in such a manner as to minimize contamination of the melt by carbon derived from the materials and the environment within the furnace. The subject furnace is constructed of carbon free materials and is housed within a conventional vacuum chamber. The furnace comprises a ceramic oxide crucible for holding the charge of metal or alloy. The heating of the crucible is achieved by a plasma-sprayed tungsten susceptor surrounding the crucible which, in turn, is heated by an RF induction coil separated from the susceptor by a cylinder of inorganic insulation. The furnace of the present invention is capable of being rapidly cycled from ambient temperatures to about 1650.degree. C. for effectively melting uranium and uranium alloys without the attendant carbon contamination problems previously encountered when using carbon-bearing furnace materials.

Non-carbon induction furnace

DOEpatents

Holcombe, C.E.; Masters, D.R.; Pfeiler, W.A.

1984-01-06

The present invention is directed to an induction furnace for melting and casting highly pure metals and alloys such as uranium and uranium alloys in such a manner as to minimize contamination of the melt by carbon derived from the materials and the environment within the furnace. The subject furnace is constructed of non-carbon materials and is housed within a conventional vacuum chamber. The furnace comprises a ceramic oxide crucible for holding the charge of metal or alloys. The heating of the crucible is achieved by a plasma-sprayed tungsten susceptor surrounding the crucible which, in turn, is heated by an rf induction coil separated from the susceptor by a cylinder of inorganic insulation. The furnace of the present invention is capable of being rapidly cycled from ambient temperatures to about 1650/sup 0/C for effectively melting uranium and uranium alloys without the attendant carbon contamination problems previously encountered when using carbon-bearing furnace materials.

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.

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.

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)

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, George F.; Vissers, Donald R.; Marshall, Simon L.; Varma, Ravi

1989-01-01

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm.sup.2 of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99.

CERAMIC FUEL ELEMENT MATERIAL FOR A NEUTRONIC REACTOR AND METHOD OF FABRICATING SAME

DOEpatents

Duckworth, W.H.

1957-12-01

This patent relates to ceramic composition, and to neutronic reactor fuel elements formed therefrom. These ceramic elements have high density and excellent strength characteristics and are formed by conventional ceramic casting and sintering at a temperature of about 2700 deg F in a nitrogen atmosphere. The composition consists of silicon carbide, silicon, uranium oxide and a very small percentage of molybdenum. Compositions containing molybdenum are markedly stronger than those lacking this ingredient.

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.

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.

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.

Synthesis of pentafluorides

DOEpatents

Asprey, L.B.; Paine, R.T. Jr.

1975-12-30

The reactions of uranium, molybdenum, rhenium, osmium and iridium hexafluorides with hydrogen gas in the presence of ultraviolet radiation or with silicon powder in an anhydrous HF slurry provide especially useful, high yield syntheses of pure pentafluorides.

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.

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.

Rolling Process Modeling Report: Finite-Element Prediction of Roll Separating Force and Rolling Defects

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

Soulami, Ayoub; Lavender, Curt A.; Paxton, Dean M.

2014-04-23

Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum (U-10Mo) alloy plate-type fuel for the U.S. high-performance research reactors. This work supports the Convert Program of the U.S. Department of Energy’s National Nuclear Security Administration (DOE/NNSA) Global Threat Reduction Initiative. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll separating forces and rolling defects. Simulations were performed using a finite-element model developed using the commercial code LS-Dyna. Simulations of the hot rolling of U-10Mo coupons encapsulated in low-carbon steel have been conducted following two different schedules. Model predictions ofmore » the roll-separation force and roll-pack thicknesses at different stages of the rolling process were compared with experimental measurements. This report discusses various attributes of the rolled coupons revealed by the model (e.g., dog-boning and thickness non-uniformity).« less

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

Creep resistant high temperature martensitic steel

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

Hawk, Jeffrey A.; Jablonski, Paul D.; Cowen, Christopher J.

The disclosure provides a creep resistant alloy having an overall composition comprised of iron, chromium, molybdenum, carbon, manganese, silicon, nickel, vanadium, niobium, nitrogen, tungsten, cobalt, tantalum, boron, and potentially additional elements. In an embodiment, the creep resistant alloy has a molybdenum equivalent Mo(eq) from 1.475 to 1.700 wt. % and a quantity (C+N) from 0.145 to 0.205. The overall composition ameliorates sources of microstructural instability such as coarsening of M.sub.23C.sub.6 carbides and MX precipitates, and mitigates or eliminates Laves and Z-phase formation. A creep resistant martensitic steel may be fabricated by preparing a melt comprised of the overall composition followedmore » by at least austenizing and tempering. The creep resistant alloy exhibits improved high-temperature creep strength in the temperature environment of around 650.degree. C.« less

Creep resistant high temperature martensitic steel

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

Hawk, Jeffrey A.; Jablonski, Paul D.; Cowen, Christopher J.

The disclosure provides a creep resistant alloy having an overall composition comprised of iron, chromium, molybdenum, carbon, manganese, silicon, nickel, vanadium, niobium, nitrogen, tungsten, cobalt, tantalum, boron, copper, and potentially additional elements. In an embodiment, the creep resistant alloy has a molybdenum equivalent Mo(eq) from 1.475 to 1.700 wt. % and a quantity (C+N) from 0.145 to 0.205. The overall composition ameliorates sources of microstructural instability such as coarsening of M.sub.23C.sub.6carbides and MX precipitates, and mitigates or eliminates Laves and Z-phase formation. A creep resistant martensitic steel may be fabricated by preparing a melt comprised of the overall composition followedmore » by at least austenizing and tempering. The creep resistant alloy exhibits improved high-temperature creep strength in the temperature environment of around 650.degree. C.« less

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.

Molybdeno-Aluminizing of Powder Metallurgy and Wrought Ti and Ti-6Al-4V alloys by Pack Cementation process

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

Tsipas, Sophia A., E-mail: stsipas@ing.uc3m.es; Go

Wear and high temperature oxidation resistance of some titanium-based alloys needs to be enhanced, and this can be effectively accomplished by surface treatment. Molybdenizing is a surface treatment where molybdenum is introduced into the surface of titanium alloys causing the formation of wear-resistant surface layers containing molybdenum, while aluminizing of titanium-based alloys has been reported to improve their high temperature oxidation properties. Whereas pack cementation and other surface modification methods have been used for molybdenizing or aluminizing of wrought and/or cast pure titanium and titanium alloys, such surface treatments have not been reported on titanium alloys produced by powder metallurgymore » (PM). Also a critical understanding of the process parameters for simultaneous one step molybdeno-aluminizing of titanium alloys by pack cementation and the predominant mechanism for this process have not been reported. The current research work describes the surface modification of titanium and Ti-6Al-4V prepared by PM by molybdeno-aluminizing and analyzes thermodynamic aspects of the deposition process. Similar coatings are also deposited to wrought Ti-6Al-4V and compared. Characterization of the coatings was carried out using scanning electron microscopy and x-ray diffraction. For both titanium and Ti-6Al-4V, the use of a powder pack containing ammonium chloride as activator leads to the deposition of molybdenum and aluminium into the surface but also introduces nitrogen causing the formation of a thin titanium nitride layer. In addition, various titanium aluminides and mixed titanium aluminium nitrides are formed. The appropriate conditions for molybdeno-aluminizing as well as the phases expected to be formed were successfully determined by thermodynamic equilibrium calculations. - Highlights: •Simultaneous co-deposition of Mo-Al onto powder metallurgy and wrought Ti alloy •Thermodynamic calculations were used to optimize deposition conditions •External TiN and internal a Mo-rich layer on all alloy substrates •Titanium aluminides and Ti-Al mixed nitrides are formed on Ti-6Al-4V •The presence of Al and V alloying elements modifies the diffusion of Mo.« less

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.

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.

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.

Molybdate Coatings for Protecting Aluminum Against Corrosion

NASA Technical Reports Server (NTRS)

Calle, Luz Marina; MacDowell, Louis G.

2005-01-01

Conversion coatings that comprise mixtures of molybdates and several additives have been subjected to a variety of tests to evaluate their effectiveness in protecting aluminum and alloys of aluminum against corrosion. Molybdate conversion coatings are under consideration as replacements for chromate conversion coatings, which have been used for more than 70 years. The chromate coatings are highly effective in protecting aluminum and its alloys against corrosion but are also toxic and carcinogenic. Hexavalent molybdenum and, hence, molybdates containing hexavalent molybdenum, have received attention recently as replacements for chromates because molybdates mimic chromates in a variety of applications but exhibit significantly lower toxicity. The tests were performed on six proprietary formulations of molybdate conversion coatings, denoted formulations A through F, on panels of aluminum alloy 2024-T3. A bare alloy panel was also included in the tests. The tests included electrochemical impedance spectroscopy (EIS), measurements of corrosion potentials, scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS), and x-ray photoelectron spectroscopy (XPS).

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)

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

REDUCTION OF INORGANIC COMPOUNDS WITH MOLECULAR HYDROGEN BY MICROCOCCUS LACTILYTICUS I.

PubMed Central

Woolfolk, C. A.; Whiteley, H. R.

1962-01-01

Woolfolk, C. A. (University of Washington, Seattle) and H. R. Whiteley. Reduction of inorganic compounds with molecular hydrogen by Micrococcus lactilyticus. I. Stoichiometry with compounds of arsenic, selenium, tellurium, transition and other elements. J. Bacteriol. 84:647–658. 1962.—Extracts of Micrococcus lactilyticus (Veillonella alcalescens) oxidize molecular hydrogen at the expense of certain compounds of arsenic, bismuth, selenium, tellurium, lead, thallium, vanadium, manganese, iron, copper, molybdenum, tungsten, osmium, ruthenium, gold, silver, and uranium, as well as molecular oxygen. Chemical and manometric data indicate that the following reductions are essentially quantitative: arsenate to arsenite, pentavalent and trivalent bismuth to the free element, selenite via elemental selenium to selenide, tellurate and tellurite to tellurium, lead dioxide and manganese dioxide to the divalent state, ferric to ferrous iron, osmium tetroxide to osmate ion, osmium dioxide and trivalent osmium to the metal, uranyl uranium to the tetravalent state, vanadate to the level of vanadyl, and polymolybdate ions to molybdenum blues with an average valence for molybdenum of +5. The results of a study of certain other hydrogenase-containing bacteria with respect to their ability to carry out some of the same reactions are also presented. PMID:14001842

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, G.F.; Vissers, D.R.; Marshall, S.L.; Varma, R.

1987-10-26

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm/sup 2/ of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99. 2 figs.

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.

40 CFR 471.71 - Effluent limitations representing the degree of effluent reduction attainable by the application...

Code of Federal Regulations, 2012 CFR

2012-07-01

... SOURCE CATEGORY Uranium Forming Subcategory § 471.71 Effluent limitations representing the degree of... off-pounds) of uranium extruded Cadium 0.117 0.052 Chromium 0.152 0.062 Copper 0.654 0.344 Lead 0.145 0.069 Nickel 0.661 0.437 Fluoride 20.5 9.08 Molybdenum 2.28 1.18 Oil and grease 6.88 4.13 TSS 14.1 6...

40 CFR 471.71 - Effluent limitations representing the degree of effluent reduction attainable by the application...

Code of Federal Regulations, 2014 CFR

2014-07-01

... SOURCE CATEGORY Uranium Forming Subcategory § 471.71 Effluent limitations representing the degree of... off-pounds) of uranium extruded Cadium 0.117 0.052 Chromium 0.152 0.062 Copper 0.654 0.344 Lead 0.145 0.069 Nickel 0.661 0.437 Fluoride 20.5 9.08 Molybdenum 2.28 1.18 Oil and grease 6.88 4.13 TSS 14.1 6...

40 CFR 471.71 - Effluent limitations representing the degree of effluent reduction attainable by the application...

Code of Federal Regulations, 2013 CFR

2013-07-01

... SOURCE CATEGORY Uranium Forming Subcategory § 471.71 Effluent limitations representing the degree of... off-pounds) of uranium extruded Cadium 0.117 0.052 Chromium 0.152 0.062 Copper 0.654 0.344 Lead 0.145 0.069 Nickel 0.661 0.437 Fluoride 20.5 9.08 Molybdenum 2.28 1.18 Oil and grease 6.88 4.13 TSS 14.1 6...

WHETSTONE ROADLESS AREA, ARIZONA.

USGS Publications Warehouse

Wrucke, Chester T.; McColly, Robert A.

1984-01-01

A mineral survey conducted has shown that areas in and adjacent to the Whetstone Roadless Area, Arizona have a substantiated resource potential for copper, lead, gold, silver, and quartz, and a probable mineral-resource potential for copper silver, lead, gold, molybdenum, tungsten, uranium, and gypsum. Copper and silver occur in a small vein deposit in the southwestern part of the roadless area. Copper, lead, silver, gold, and molybdenum are known in veins associated with a porphyry copper deposit in a reentrant near the southern border of the roadless area. Vein deposits of tungsten and uranium are possible in the northeast part of the roadless area near areas of known production of these commodities. Demonstrated resources of quartz for smelter flux extend into the roadless area from the Ricketts mine. Areas of probable potential for gypsum resources also occur within the roadless area. No potential for fossil fuel resources was identified in the study.

Corrosion resistant iron aluminides exhibiting improved mechanical properties and corrosion resistance

DOEpatents

Liu, C.T.; McKamey, C.G.; Tortorelli, P.F.; David, S.A.

1994-06-14

The specification discloses a corrosion-resistant intermetallic alloy comprising, in atomic percent, an FeAl iron aluminide containing from about 30 to about 40% aluminum alloyed with from about 0.01 to 0.4% zirconium and from 0.01 to about 0.8% boron. The alloy exhibits considerably improved room temperature ductility for enhanced usefulness in structural applications. The high temperature strength and fabricability is improved by alloying with molybdenum, carbon, chromium and vanadium. 9 figs.

Corrosion resistant iron aluminides exhibiting improved mechanical properties and corrosion resistance

DOEpatents

Liu, Chain T.; McKamey, Claudette G.; Tortorelli, Peter F.; David, Stan A.

1994-01-01

The specification discloses a corrosion-resistant intermetallic alloy comprising, in atomic percent, an FeAl iron aluminide containing from about 30 to about 40% aluminum alloyed with from about 0.01 to 0.4% zirconium and from 0.01 to about 0.8% boron. The alloy exhibits considerably improved room temperature ductility for enhanced usefulness in structural applications. The high temperature strength and fabricability is improved by alloying with molybdenum, carbon, chromium and vanadium.

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.

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.

Data on Streamflow and Quality of Water and Bottom Sediment in and near Humboldt Wildlife Management Area, Churchill and Pershing Counties, Nevada, 1998-2000

USGS Publications Warehouse

Paul, Angela P.; Thodal, Carl E.

2003-01-01

This study was initiated to expand upon previous findings that indicated concentrations of dissolved solids, arsenic, boron, mercury, molybdenum, selenium, and uranium were either above geochemical background concentrations or were approaching or exceeding ecological criteria in the lower Humboldt River system. Data were collected from May 1998 to September 2000 to further characterize streamflow and surface-water and bottom-sediment quality in the lower Humboldt River, selected agricultural drains, Upper Humboldt Lake, and Lower Humboldt Drain (ephemeral outflow from Humboldt Sink). During this study, flow in the lower Humboldt River was either at or above average. Flows in Army and Toulon Drains generally were higher than reported in previous investigations. An unnamed agricultural drain contributed a small amount to the flow measured in Army Drain. In general, measured concentrations of sodium, chloride, dissolved solids, arsenic, boron, molybdenum, and uranium were higher in water from agricultural drains than in Humboldt River water during this study. Mercury concentrations in water samples collected during the study period typically were below the laboratory reporting level. However, low-level mercury analyses showed that samples collected in August 1999 from Army Drain had higher mercury concentrations than those collected from the river or Toulon Drain or the Lower Humboldt Drain. Ecological criteria and effect concentrations for sodium, chloride, dissolved solids, arsenic, boron, mercury, and molybdenum were exceeded in some water samples collected as part of this study. Although water samples from the agricultural drains typically contained higher concentrations of sodium, chloride, dissolved solids, arsenic, boron, and uranium, greater instantaneous loads of these constituents were carried in the river near Lovelock than in agricultural drains during periods of high flow or non-irrigation. During this study, the high flows in the lower Humboldt River produced the maximum instantaneous loads of sodium, chloride, dissolved solids, arsenic, boron, molybdenum, and uranium at all river-sampling sites, except molybdenum near Imlay. Nevada Division of Environmental Protection monitoring reports on mine-dewatering discharge for permitted releases of treated effluent to the surface waters of the Humboldt River and its tributaries were reviewed for reported discharges and trace-element concentrations from June 1998 to September 1999. These data were compared with similar information for the river near Imlay. In all bottom sediments collected for this study, arsenic concentrations exceeded the Canadian Freshwater Interim Sediment-Quality Guideline for the protection of aquatic life and probable-effect level (concentration). Sediments collected near Imlay, Rye Patch Reservoir, Lovelock, and from Toulon Drain and Army Drain were found to contain cadmium and chromium concentrations that exceeded Canadian criteria. Chromium concentrations in sediments collected from these sites also exceeded the consensus-based threshold-effect concentration. The Canadian criterion for sediment copper concentration was exceeded in sediments collected from the Humboldt River near Lovelock and from Toulon, Army, and the unnamed agricultural drains. Mercury in sediments collected near Imlay and from Toulon Drain in August 1999 exceeded the U.S. Department of the Interior sediment probable-effect level. Nickel concentrations in sediments collected during this study were above the consensus-based threshold-effect concentration. All other river and drain sediments had constituent concentrations below protective criteria and toxicity thresholds. In Upper Humboldt Lake, chloride, dissolved solids, arsenic, boron, molybdenum, and uranium concentrations in surface-water samples collected near the mouth of the Humboldt River generally were higher than in samples collected near the mouth of Army Drain. Ecological criteria or effect con

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

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.

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

Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºCmore » to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant 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 Undergraduate Thesis by William Sames, Research Fellow, entitled “Uranium Metal Powder Production, Particle Distribution Analysis, and Reaction Rate Studies of a Hydride-Dehydride Process"« less

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

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

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.

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.

40 CFR 471.71 - Effluent limitations representing the degree of effluent reduction attainable by the application...

Code of Federal Regulations, 2010 CFR

2010-07-01

... CATEGORY Uranium Forming Subcategory § 471.71 Effluent limitations representing the degree of effluent... uranium extruded Cadium 0.117 0.052 Chromium 0.152 0.062 Copper 0.654 0.344 Lead 0.145 0.069 Nickel 0.661 0.437 Fluoride 20.5 9.08 Molybdenum 2.28 1.18 Oil and grease 6.88 4.13 TSS 14.1 6.71 pH (1) (1) 1...

40 CFR 471.71 - Effluent limitations representing the degree of effluent reduction attainable by the application...

Code of Federal Regulations, 2011 CFR

2011-07-01

... CATEGORY Uranium Forming Subcategory § 471.71 Effluent limitations representing the degree of effluent... uranium extruded Cadium 0.117 0.052 Chromium 0.152 0.062 Copper 0.654 0.344 Lead 0.145 0.069 Nickel 0.661 0.437 Fluoride 20.5 9.08 Molybdenum 2.28 1.18 Oil and grease 6.88 4.13 TSS 14.1 6.71 pH (1) (1) 1...

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

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.

Surface compositional variations of Mo-47Re alloy as a function of temperature

NASA Technical Reports Server (NTRS)

Hoekje, S. J.; Outlaw, R. A.; Sankaran, S. N.

1993-01-01

Molybdenum-rhenium alloys are candidate materials for the National Aero-Space Plane (NASP) as well as for other applications in generic hypersonics. These materials are expected to be subjected to high-temperature (above 1200 C) casual hydrogen (below 50 torr), which could potentially degrade the material strength. Since the uptake of hydrogen may be controlled by the contaminant surface barriers, a study of Mo-47Re was conducted to examine the variations in surface composition as a function of temperature from 25 C to 1000 C. Pure molybdenum and rhenium were also examined and the results compared with those for the alloy. The analytical techniques employed were Auger electron spectroscopy, electron energy loss spectroscopy, ion scattering spectroscopy, and x ray photoelectron spectroscopy. The native surface was rich in metallic oxides that disappeared at elevated temperatures. As the temperature increased, the carbon and oxygen disappeared by 800 C and the surface was subsequently populated by the segregation of silicon, presumably from the grain boundaries. The alloy readily chemisorbed oxygen, which disappeared with heating. The disappearance temperature progressively increased for successive dosings. When the alloy was exposed to 800 torr of hydrogen at 900 C for 1 hour, no hydrogen interaction was observed.

High-temperature fabricable nickel-iron aluminides

DOEpatents

Liu, Chain T.

1988-02-02

Nickel-iron aluminides are described that are based on Ni.sub.3 Al, and have significant iron content, to which additions of hafnium, boron, carbon and cerium are made resulting in Ni.sub.3 Al base alloys that can be fabricated at higher temperatures than similar alloys previously developed. Further addition of molybdenum improves oxidation and cracking resistance. These alloys possess the advantages of ductility, hot fabricability, strength, and oxidation resistance.

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.

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)

Enhancements to High Temperature In-Pile Thermocouple Performance

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

J.C. Crepeau; J.L. Rempe; J.E. Daw

2008-03-31

A joint University of Idaho (UI) and Idaho National Laboratory (INL) University Nuclear Research Initiative (UNERI) was to initiated to extend initial INL efforts to develop doped molybdenum/niobium alloy High Temperature Irradiation Resistant Thermocouples (HTIR-TCs). The overall objective of this UNERI was to develop recommendations for an optimized thermocouple design for high temperature, long duration, in-pile testing by expanding upon results from initial INL efforts. Tasks to quantify the impact of candidate enhancements, such as alternate alloys, alternate geometries, and alternate thermocouple fabrication techniques, on thermocouple performance were completed at INL's High Temperature Test Laboratory (HTTL), a state of themore » art facility equipped with specialized equipment and trained staff in the area of high temperature instrumentation development and evaluation. Key results of these evaluations, which are documented in this report, are as follows. The doped molybdenum and Nb-1%Zr, which were proposed in the initial INL HTIR-TC design, were found to retain ductility better than the developmental molybdenum-low niobium alloys and the niobium-low molybdenum alloys evaluated. Hence, the performance and lower cost of the commercially available KW-Mo makes a thermocouple containing KW-Mo and Nb-1%Zr the best option at this time. HTIR-TCs containing larger diameter wires offer the potential to increase HTIR-TC stability and reliability at higher temperatures. HTIR-TC heat treatment temperatures and times should be limited to not more than 100 C above the proposed operating temperatures and to durations of at least 4 to 5 hours. Preliminary investigations suggest that the performance of swaged and loose assembly HTIR-TC designs is similar. However, the swaged designs are less expensive and easier to construct. In addition to optimizing HTIR-TC performance, This UNERI project provided unique opportunities to several University of Idaho students, allowing them to become familiar with the techniques and equipment used for specialized high temperature instrumentation fabrication and evaluation and to author/coauthor several key conference papers and journal articles.« less

Fuel preparation for use in the production of medical isotopes

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

Policke, Timothy A.; Aase, Scott B.; Stagg, William R.

The present invention relates generally to the field of medical isotope production by fission of uranium-235 and the fuel utilized therein (e.g., the production of suitable Low Enriched Uranium (LEU is uranium having 20 weight percent or less uranium-235) fuel for medical isotope production) and, in particular to a method for producing LEU fuel and a LEU fuel product that is suitable for use in the production of medical isotopes. In one embodiment, the LEU fuel of the present invention is designed to be utilized in an Aqueous Homogeneous Reactor (AHR) for the production of various medical isotopes including, butmore » not limited to, molybdenum-99, cesium-137, iodine-131, strontium-89, xenon-133 and yttrium-90.« less

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.

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)

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

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

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

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

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated inmore » the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.« less

Geochemistry and migration of contaminants at the Weldon Spring chemical plant site, St. Charles County, Missouri, 1989-91

USGS Publications Warehouse

Schumacher, John G.

1993-01-01

The geochemistry of the shallow aquifer and geochemical controls on the migration of uranium and other constituents from raffinate pits were determined at the Weldon Spring chemical plant site. Surface-water samples from the raffinate pits con- tained large concentrations of calcium, magnesium, sodium, potassium, sulfate, nitrite, lithium, moly- bdenum, strontium, vanadium, and uranium. Analyses of interstitial-water samples from raffinate pit 3 indicated that concentrations of most constituents increased with increasing depth below the water- sediment interface. Nitrate and uranium were not chemically reduced and attenuated within the raffinate pits and can be expected to migrate into the overburden. Laboratory sorption experiments were performed to evaluate the effect of pH value on the sorption of several raffinate constituents by the overburden. No sorption of calcium, sodium, sulfate, nitrate, or lithium was observed. Sorption of molybdenum was dependent on solution pH and sorption of uranium was dependent on solution pH and carbonate concentration. The sorption of uranium and molybdenum was consistent with sorption controlled by oxyhydroxides. The quality of water collected in overburden lysimeters near raffinate pit 4 can be modeled as a mixture of water from raffinate pits 3 and 4, and an uncontaminated com- ponent in a system at equilibrium with ferrihydrite and calcite. Increased constituent concentrations in a perennial spring north of the site were the result of a subsurface connection between the spring and several losing stream segments receiving runoff from the site, in addition to seepage from the raffinate pits.

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)

Resource potential for commodities in addition to Uranium in sandstone-hosted deposits: Chapter 13

USGS Publications Warehouse

Breit, George N.

2016-01-01

Sandstone-hosted deposits mined primarily for their uranium content also have been a source of vanadium and modest amounts of copper. Processing of these ores has also recovered small amounts of molybdenum, rhenium, rare earth elements, scandium, and selenium. These deposits share a generally common origin, but variations in the source of metals, composition of ore-forming solutions, and geologic history result in complex variability in deposit composition. This heterogeneity is evident regionally within the same host rock, as well as within districts. Future recovery of elements associated with uranium in these deposits will be strongly dependent on mining and ore-processing methods.

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.

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.

Materials Research in Support of Superconducting Machinery V

DTIC Science & Technology

1976-04-01

GTAW , EB, GMAW), brazing, and soldering from 4-300 K. Properties include tensile, notched tensile, fracture toughness, and fatigue crack growth...include: aluminum alloys 1100, 2014, 2219; a nicke1- chromium -iron alloy; iron-47.5 nickel; and the composite materials boron/aluminum, boron/epoxy, S...nickel" by H. M. Ledbetter and D. T. Read. (3) N. jkel- chromium -iron-molybdenum alloy. There is an accompanying reprint of our previously described

Stress Corrosion Cracking and Hydrogen Embrittlement of Thick Section High Strength Low Alloy Steel

DTIC Science & Technology

1986-06-01

copper and especially molybdenum. Dual phase HSLA steels are comprised of islands of martensite or bainite in a ferrite matrix. The... Copper Steels", TransactionN AIME, Volume 105, pp. 133-166, 1933. 60. Creswick, W. E., "Commercial Development of a Rimmed Low Alloy Precipitation ... precipitates all serve to minimize the aggregate effects of hydrogen. 82 - ------- ------ - 3. MATERIAL 3.1 bSLA STEELS High strength low alloy

Characterization of microstructural, mechanical and thermophysical properties of Th-52U alloy

NASA Astrophysics Data System (ADS)

Das, Santanu; Kaity, S.; Kumar, R.; Banerjee, J.; Roy, S. B.; Chaudhari, G. P.; Daniel, B. S. S.

2016-11-01

Th-52 wt.% U alloy has a microstructure featuring interspersed networks of uranium rich and thorium rich phases. Room temperature hardness of the alloy is more than twice that of unalloyed thorium. The alloy age hardens (550 °C) only slightly (peak hardness/hardness of solution heated and quenched = 1.05). Room temperature thermal conductivity (25.6 W m-1 °C-1) is close to that of uranium and most of the binary and ternary metallic alloy fuel materials. Average linear coefficient of thermal expansion (CTE) of Th-52 wt.% U alloy [11.2 × 10-06 °C-1 (27-290 °C) and 16.75 × 10-06 °C-1 (27-600 °C)] are comparable with that of many metallic alloy fuel candidates. Th-52 wt.% U alloy with non-age hardenable microstructure, appreciable thermal conductivity, moderate thermal expansion may find metallic fuel applications in nuclear reactors.

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.

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.

A New Approach of Designing Superalloys for Low Density

NASA Technical Reports Server (NTRS)

MacKay, Rebecca A.; Gabb, Timothy P.; Smialek, James L.; Nathal, Michael V.

2010-01-01

New low-density single-crystal (LDS) alloy, have bee. developed for turbine blade applications, which have the potential for significant improvements in the thrust-to-weight ratio over current production superalloys. An innovative alloying strategy was wed to achieve alloy density reductions, high-temperature creep resistance, microstructural stability, and cyclic oxidation resistance. The alloy design relies on molybdenum as a potent. lower-density solid-solution strengthener in the nickel-based superalloy. Low alloy density was also achieved with modest rhenium levels tmd the absence of tungsten. Microstructural, physical mechanical, and environmental testing demonstrated the feasibility of this new LDS superalloy design.

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

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.

Hazard evaluation of inorganics, singly and in mixtures, to Flannelmouth Sucker Catostomus latipinnis in the San Juan River, New Mexico

USGS Publications Warehouse

Hamilton, S.J.; Buhl, K.J.

1997-01-01

Larval flannelmouth sucker (Catostomus latipinnis) were exposed to arsenate, boron, copper, molybdenum, selenate, selenite, uranium, vanadium, and zinc singly, and to five mixtures of five to nine inorganics. The exposures were conducted in reconstituted water representative of the San Juan River near Shiprock, New Mexico. The mixtures simulated environmental ratios reported for sites along the San Juan River (San Juan River backwater, Fruitland marsh, Hogback East Drain, Mancos River, and McElmo Creek). The rank order of the individual inorganics, from most to least toxic, was: copper > zinc > vanadium > selenite > selenate > arsenate > uranium > boron > molybdenum. All five mixtures exhibited additive toxicity to flannelmouth sucker. In a limited number of tests, 44-day-old and 13-day-old larvae exhibited no difference in sensitivity to three mixtures. Copper was the major toxic component in four mixtures (San Juan backwater, Hogback East Drain, Mancos River, and McElmo Creek), whereas zinc was the major toxic component in the Fruitland marsh mixture, which did not contain copper. The Hogback East Drain was the most toxic mixture tested. Comparison of 96-h LC50values with reported environmental water concentrations from the San Juan River revealed low hazard ratios for arsenic, boron, molybdenum, selenate, selenite, uranium, and vanadium, moderate hazard ratios for zinc and the Fruitland marsh mixture, and high hazard ratios for copper at three sites and four environmental mixtures representing a San Juan backwater, Hogback East Drain, Mancos River, and McElmo Creek. The high hazard ratios suggest that inorganic contaminants could adversely affect larval flannelmouth sucker in the San Juan River at four sites receiving elevated inorganics.

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.

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.

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

Comparison of abundances of chemical elements in mineralized and unmineralized sandstone of the Brushy Basin Member of the Morrison Formation, Smith Lake District, Grants uranium region, New Mexico

USGS Publications Warehouse

Pierson, C.T.; Spirakis, C.S.; Robertson, J.F.

1983-01-01

Statistical treatment of analytical data from the Mariano Lake and Ruby uranium deposits in the Smith Lake district, New Mexico, indicates that organic carbon, arsenic, barium, calcium, cobalt, copper, gallium, iron, lead, manganese, molybdenum, nickel, selenium, strontium, sulfur, vanadium, yttrium, and zirconium are concentrated along with uranium in primary ore. Comparison of the Smith Lake data with information from other primary deposits in the Grants uranium region and elsewhere in the Morrison Formation of the Colorado Plateau suggests that these elements, with the possible exceptions of zirconium and gallium and with the probable addition of aluminum and magnesium, are typically associated with primary, tabular uranium deposits. Chemical differences between the Ruby and Mariano Lake deposits are consistent with the interpretation that the Ruby deposit has been more affected by post-mineralization oxidizing solutions than has the Mariano Lake deposit.

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.

ERDA, RBS, TEM and SEM characterization of microstructural evolution in helium-implanted Hastelloy N alloy

NASA Astrophysics Data System (ADS)

Gao, Jie; Bao, Liangman; Huang, Hefei; Li, Yan; Lei, Qiantao; Deng, Qi; Liu, Zhe; Yang, Guo; Shi, Liqun

2017-05-01

Hastelloy N alloy was implanted with 30 keV, 5 × 1016 ions/cm2 helium ions at room temperature, and subsequent annealed at 600 °C for 1 h and further annealed at 850 °C for 5 h in vacuum. Using elastic recoil detection analysis (ERDA) and transmission electron microscopy (TEM), the depth profiles of helium concentration and helium bubbles in helium-implanted Hastelloy N alloy were investigated, respectively. The diffusion of helium and molybdenum elements to surface occurred during the vacuum annealing at 850 °C (5 h). It was also observed that bubbles in molybdenum-enriched region were much larger in size than those in deeper region. In addition, it is worth noting that plenty of nano-holes can be observed on the surface of helium-implanted sample after high temperature annealing by scanning electron microscope (SEM). This observation provides the evidence for the occurrence of helium release, which can be also inferred from the results of ERDA and TEM analysis.

Influence of Chemical Composition on Rupture Properties at 1200 Degrees F. of Forged Chromium-Cobalt-Nickel-Iron Base Alloys in Solution-Treated and Aged Condition

NASA Technical Reports Server (NTRS)

Reynolds, E E; Freeman, J W; White, A E

1951-01-01

The influence of systematic variations of chemical composition on rupture properties at 1200 degrees F. was determined for 62 modifications of a basic alloy containing 20 percent chromium, 20 percent nickel, 20 percent cobalt, 3 percent molybdenum, 2 percent tungsten, 1 percent columbium, 0.15 percent carbon, 1.7 percent manganese, 0.5 percent silicon, 0.12 percent nitrogen and the balance iron. These modifications included individual variations of each of 10 elements present and simultaneous variations of molybdenum, tungsten, and columbium. Laboratory induction furnace heats were hot-forged to round bar stock, solution-treated at 2200 degrees F., and aged at 1400 degrees F. The melting and fabrication conditions were carefully controlled in order to minimize all variable effects on properties except chemical composition. Information is presented which indicates that melting and hot-working conditions play an important role in high-temperature properties of alloys of the type investigated.

Phase and crystallite size analysis of (Ti1-xMox)C-(Ni,Cr) cermet obtained by mechanical alloying

NASA Astrophysics Data System (ADS)

Suryana, Anis, Muhammad; Manaf, Azwar

2018-04-01

In this paper, we report the phase and crystallite size analysis of (Ti1-xMox)C-(Ni,Cr) with x = 0-0.5 cermet obtained by mechanical alloying of Ti, Mo, Ni, Cr and C elemental powders using a high-energy shaker ball mill under wet condition for 10 hours. The process used toluene as process control agent and the ball to mass ratio was 10:1. The mechanically milled powder was then consolidated and subsequently heated at a temperature 850°C for 2 hours under an argon flow to prevent oxidation. The product was characterized by X-ray diffraction (XRD) and scanning electron microscope equipped with energy dispersive analyzer. Results shown that, by the selection of appropriate condition during the mechanical alloying process, a metastable Ti-Ni-Cr-C powders could be obtained. The powder then allowed the in situ synthesis of TiC-(Ni,Cr) cermet which took place during exposure time at a high temperature that applied in reactive sintering step. Addition to molybdenum has caused shifting the TiC XRD peaks to a slightly higher angle which indicated that molybdenum dissolved in TiC phase. The crystallite size distribution of TiC is discussed in the report, which showing that the mean size decreased with the addition of molybdenum.

Losses and depolarization of ultracold neutrons on neutron guide and storage materials

NASA Astrophysics Data System (ADS)

Bondar, V.; Chesnevskaya, S.; Daum, M.; Franke, B.; Geltenbort, P.; Göltl, L.; Gutsmiedl, E.; Karch, J.; Kasprzak, M.; Kessler, G.; Kirch, K.; Koch, H.-C.; Kraft, A.; Lauer, T.; Lauss, B.; Pierre, E.; Pignol, G.; Reggiani, D.; Schmidt-Wellenburg, P.; Sobolev, Yu.; Zechlau, T.; Zsigmond, G.

2017-09-01

At Institut Laue-Langevin (ILL) and Paul Scherrer Institute (PSI), we have measured the losses and depolarization probabilities of ultracold neutrons on various materials: (i) nickel-molybdenum alloys with weight percentages of 82/18, 85/15, 88/12, 91/9, and 94/6 and natural nickel Ni100, (ii) nickel-vanadium NiV93/7, (iii) copper, and (iv) deuterated polystyrene (dPS). For the different samples, storage-time constants up to ˜460 s were obtained at room temperature. The corresponding loss parameters for ultracold neutrons, η , varied between 1.0 ×10-4 and 2.2 ×10-4 . All η values are in agreement with theory except for dPS, where anomalous losses at room temperature were established with four standard deviations. The depolarization probabilities per wall collision β measured with unprecedented sensitivity varied between 0.7 ×10-6 and 9.0 ×10-6 . Our depolarization result for copper differs from other experiments by 4.4 and 15.8 standard deviations. The β values of the paramagnetic NiMo alloys over molybdenum content show an increase of β with increasing Mo content. This is in disagreement with expectations from literature. Finally, ferromagnetic behavior of NiMo alloys at room temperature was found for molybdenum contents of 6.5 at.% or less and paramagnetic behavior for more than 8.7 at.%. This may contribute to solving an ambiguity in literature.

Effect of composition and austenite deformation on the transformation characteristics of low-carbon and ultralow-carbon microalloyed steels

NASA Astrophysics Data System (ADS)

Cizek, P.; Wynne, B. P.; Davies, C. H. J.; Muddle, B. C.; Hodgson, P. D.

2002-05-01

Deformation dilatometry has been used to simulate controlled hot rolling followed by controlled cooling of a group of low- and ultralow-carbon microalloyed steels containing additions of boron and/or molybdenum to enhance hardenability. Each alloy was subjected to simulated recrystallization and nonrecrystallization rolling schedules, followed by controlled cooling at rates from 0.1 °C/s to about 100 °C/s, and the corresponding continuous-cooling-transformation (CCT) diagrams were constructed. The resultant microstructures ranged from polygonal ferrite (PF) for combinations of slow cooling rates and low alloying element contents, through to bainitic ferrite accompanied by martensite for fast cooling rates and high concentrations of alloying elements. Combined additions of boron and molybdenum were found to be most effective in increasing steel hardenability, while boron was significantly more effective than molybdenum as a single addition, especially at the ultralow carbon content. Severe plastic deformation of the parent austenite (>0.45) markedly enhanced PF formation in those steels in which this microstructural constituent was formed, indicating a significant effective decrease in their hardenability. In contrast, in those steels in which only nonequilibrium ferrite microstructures were formed, the decreases in hardenability were relatively small, reflecting the lack of sensitivity to strain in the austenite of those microstructural constituents forming in the absence of PF.

BRAZE BONDING OF COLUMBIUM

DOEpatents

Heestand, R.L.; Picklesimer, M.L.

1962-07-31

A method of brazing niobium parts together is described. The surfaces of the parts to be brazed together are placed in abutting relationship with a brazing alloy disposed adjacent. The alloy consists essentially of, by weight, 12 to 25% niobium, 0.5 to 5% molybdenum, and the balance zirconium, The alloy is heated to at least its melting point to braze the parts together. The brazed joint is then cooled. The heating, melting and cooling take place in an inert atmosphere. (AEC)

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.

A review of refractory materials for vapor-anode AMTEC cells

NASA Astrophysics Data System (ADS)

King, Jeffrey C.; El-Genk, M. S.

2000-01-01

Recently, refractory alloys have been considered as structural materials for vapor-anode Alkali Metal Thermal-to-Electric Conversion (AMTEC) cells, for extended (7-15 years) space missions. This paper reviewed the existing database for refractory metals and alloys of potential use as structural materials for vapor-anode sodium AMTEC cells. In addition to requiring that the vapor pressure of the material be below 10-9 torr (133 nPa) at a typical hot side temperature of 1200 K, other screening considerations were: (a) low thermal conductivity, low thermal radiation emissivity, and low linear thermal expansion coefficient; (b) low ductile-to-brittle transition temperature, high yield and rupture strengths and high strength-to-density ratio; and (c) good compatibility with the sodium AMTEC operating environment, including high corrosion resistance to sodium in both the liquid and vapor phases. Nb-1Zr (niobium-1% zirconium) alloy is recommended for the hot end structures of the cell. The niobium alloy C-103, which contains the oxygen gettering elements zirconium and hafnium as well as titanium, is recommended for the colder cell structure. This alloy is stronger and less thermally conductive than Nb-1Zr, and its use in the cell wall reduces parasitic heat losses by conduction to the condenser. The molybdenum alloy Mo-44.5Re (molybdenum-44.5% rhenium) is also recommended as a possible alternative for both structures if known problems with oxygen pick up and embrittlement of the niobium alloys proves to be intractable. .

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.

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.

Compositions and methods for treating nuclear fuel

DOEpatents

Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

2013-08-13

Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Compositions and methods for treating nuclear fuel

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

Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K

Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Nickel base alloy. [for gas turbine engine stator vanes

NASA Technical Reports Server (NTRS)

Freche, J. C.; Waters, W. J. (Inventor)

1977-01-01

A nickel base superalloy for use at temperatures of 2000 F (1095 C) to 2200 F (1205 C) was developed for use as stator vane material in advanced gas turbine engines. The alloy has a nominal composition in weight percent of 16 tungsten, 7 aluminum, 1 molybdenum, 2 columbium, 0.3 zirconium, 0.2 carbon and the balance nickel.

The Quantitative Microstructural Characterization of Multipass TIG Ultra Low Carbon Bainitic Steel Weldments and Correlation with Mechanical Properties

DTIC Science & Technology

1993-09-01

in TIG weldments. The alloying elements used in ULCB steels are; Carbon (C), Manganese (Mn), Molybdenum (Mo), Nickel (Ni), Niobium (Nb), Chromium (Cr...process. 7 C. WELDING PROCESSES 1. Tungsten Inert Gas (TIG) Welding Tungsten Inert Gas (TIG) Welding (or Gas Tungsten Arc Welding ( GTAW )), produces... chromium (Cr), molybdenum (Mo), and sometimes vanadium (V). Reheat cracking occurs in the HAZ during postweld stress relieving, especially in thick

Shield materials recommended for space power nuclear reactors

NASA Technical Reports Server (NTRS)

Kaszubinski, L. J.

1973-01-01

Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

Special nuclear material simulation device

DOEpatents

Leckey, John H.; DeMint, Amy; Gooch, Jack; Hawk, Todd; Pickett, Chris A.; Blessinger, Chris; York, Robbie L.

2014-08-12

An apparatus for simulating special nuclear material is provided. The apparatus typically contains a small quantity of special nuclear material (SNM) in a configuration that simulates a much larger quantity of SNM. Generally the apparatus includes a spherical shell that is formed from an alloy containing a small quantity of highly enriched uranium. Also typically provided is a core of depleted uranium. A spacer, typically aluminum, may be used to separate the depleted uranium from the shell of uranium alloy. A cladding, typically made of titanium, is provided to seal the source. Methods are provided to simulate SNM for testing radiation monitoring portals. Typically the methods use at least one primary SNM spectral line and exclude at least one secondary SNM spectral line.

Thermodynamic properties of La-Ga-Al and U-Ga-Al alloys and the separation factor of U/La couple in the molten salt-liquid metal system

NASA Astrophysics Data System (ADS)

Novoselova, A.; Smolenski, V.; Volkovich, V. A.; Ivanov, A. B.; Osipenko, A.; Griffiths, T. R.

2015-11-01

The electrochemical behaviour of lanthanum and uranium was studied in fused 3LiCl-2KCl eutectic and Ga-Al eutectic liquid metal alloy between 723 and 823 K. Electrode potentials were recorded vs. Cl-/Cl2 reference electrode and the temperature dependencies of the apparent standard potentials of La-(Ga-Al) and U-(Ga-Al) alloys were determined. Lanthanum and uranium activity coefficients and U/La couple separation factor were calculated. Partial excess free Gibbs energy, partial enthalpy of mixing and partial excess entropy of La-(Ga-Al) and U-(Ga-Al) alloys were estimated.

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.

NUCLEAR FUEL COMPOSITION

DOEpatents

Spedding, F.H.; Wilhelm, H.A.

1960-05-31

A novel reactor composition for use in a self-sustaining fast nuclear reactor is described. More particularly, a fuel alloy comprising thorium and uranium-235 is de scribed, the uranium-235 existing in approximately the same amount that it is found in natural uranium, i.e., 1.4%.

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.

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.

PYROMETALLURGICAL METHOD

DOEpatents

Nelson, P.A.

1961-07-18

The liquid--liquid extraction of plutonium by magnesium from uranium or uranium--chromium alloy is described. Calcium is added to magnesium in about eutectic proportions, which results in a purer plutonium.

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)

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.

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.

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.

Computer Aided Design of Ni-Based Single Crystal Superalloy for Industrial Gas Turbine Blades

NASA Astrophysics Data System (ADS)

Wei, Xianping; Gong, Xiufang; Yang, Gongxian; Wang, Haiwei; Li, Haisong; Chen, Xueda; Gao, Zhenhuan; Xu, Yongfeng; Yang, Ming

The influence of molybdenum, tungsten and cobalt on stress-rupture properties of single crystal superalloy PWA1483 has been investigated using the simulated calculation of JMatPro software which ha s been widely used to develop single crystal superalloy, and the effect of alloying element on the stability of strengthening phase has been revealed by using the Thermo-Calc software. Those properties calculation results showed that the increasing of alloy content could facilitate the precipitation of TCP phases and increase the lattice misfit between γ and γ' phase, and the effect of molybdenum, tantalum was the strongest and that of cobalt was the weakest. Then the chemical composition was optimized, and the selected compositions showed excellent microstructure stability and stress-rupture properties by the confirmation of d-electrons concept and software calculation.

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.

Molybdenum Availability Is Key to Nitrate Removal in Contaminated Groundwater Environments

PubMed Central

Thorgersen, Michael P.; Lancaster, W. Andrew; Vaccaro, Brian J.; Poole, Farris L.; Rocha, Andrea M.; Mehlhorn, Tonia; Pettenato, Angelica; Ray, Jayashree; Waters, R. Jordan; Melnyk, Ryan A.; Chakraborty, Romy; Deutschbauer, Adam M.; Arkin, Adam P.

2015-01-01

The concentrations of molybdenum (Mo) and 25 other metals were measured in groundwater samples from 80 wells on the Oak Ridge Reservation (ORR) (Oak Ridge, TN), many of which are contaminated with nitrate, as well as uranium and various other metals. The concentrations of nitrate and uranium were in the ranges of 0.1 μM to 230 mM and <0.2 nM to 580 μM, respectively. Almost all metals examined had significantly greater median concentrations in a subset of wells that were highly contaminated with uranium (≥126 nM). They included cadmium, manganese, and cobalt, which were 1,300- to 2,700-fold higher. A notable exception, however, was Mo, which had a lower median concentration in the uranium-contaminated wells. This is significant, because Mo is essential in the dissimilatory nitrate reduction branch of the global nitrogen cycle. It is required at the catalytic site of nitrate reductase, the enzyme that reduces nitrate to nitrite. Moreover, more than 85% of the groundwater samples contained less than 10 nM Mo, whereas concentrations of 10 to 100 nM Mo were required for efficient growth by nitrate reduction for two Pseudomonas strains isolated from ORR wells and by a model denitrifier, Pseudomonas stutzeri RCH2. Higher concentrations of Mo tended to inhibit the growth of these strains due to the accumulation of toxic concentrations of nitrite, and this effect was exacerbated at high nitrate concentrations. The relevance of these results to a Mo-based nitrate removal strategy and the potential community-driving role that Mo plays in contaminated environments are discussed. PMID:25979890

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.

Ground-water contamination near a uranium tailings disposal site in Colorado

USGS Publications Warehouse

Goode, Daniel J.; Wilder, Russell J.

1987-01-01

Contaminants from uranium tailings disposed of at an active mill in Colorado have seeped into the shallow ground water onsite. This ground water discharges into the Arkansas River Valley through a superposed stream channel cut in the resistant sandstone ridge at the edge of a synclinal basin. In the river valley, seasonal surface-water irrigation has a significant impact on hydrodynamics. Water levels in residential wells fluctuate up to 20 ft and concentrations of uranium, molybdenum, and other contaminants also vary seasonally, with highest concentrations in the Spring, prior to irrigation, and lowest concentrations in the Fall. Results of a simple transient mixing cell model support the hypothesis that lateral ground-water inflow, and not irrigation recharge, is the source of ground-water contamination.

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

Joshi, Vineet V.; Paxton, Dean M.; Lavender, Curt A.

Over the past several years Pacific Northwest National Laboratory (PNNL) has been actively involved in supporting the U.S. Department of Energy National Nuclear Security Administration Office of Material Management and Minimization (formerly Global Threat Reduction Initiative). The U.S. High- Power Research Reactor (USHPRR) project is developing alternatives to existing highly enriched uranium alloy fuel to reduce the proliferation threat. One option for a high-density metal fuel is uranium alloyed with 10 wt% molybdenum (U-10Mo). Forming the U-10Mo fuel plates/foils via rolling is an effective technique and is actively being pursued as part of the baseline manufacturing process. The processing ofmore » these fuel plates requires systematic investigation/understanding of the pre- and post-rolling microstructure, end-state mechanical properties, residual stresses, and defects, their effect on the mill during processing, and eventually, their in-reactor performance. In the work documented herein, studies were conducted to determine the effect of cold and hot rolling the as-cast and homogenized U-10Mo on its microstructure and hardness. The samples were homogenized at 900°C for 48 h, then later annealed for several durations and temperatures to investigate the effect on the material’s microstructure and hardness. The rolling of the as-cast plate, both hot and cold, was observed to form a molybdenum-rich and -lean banded structure. The cold rolling was ineffective, and in some cases exacerbated the as-cast defects. The grains elongated along the rolling direction and formed a pancake shape, while the carbides fractured perpendicularly to the rolling direction and left porosity between fractured particles of UC. The subsequent annealing of these samples at sub-eutectoid temperatures led to rapid precipitation of the ' lamellar phase, mainly in the molybdenum-lean regions. Annealing the samples above the eutectoid temperature did not refine the grain size or the banded microstructure. However, annealing the samples led to quick recovery in hardness as evidenced by a drop in Vickers hardness of 20%. Hot rolling was performed at 650 and 800°C. The hot-rolling mill loads (load separation force) were approximately 40 to 50% less than the cold-rolling for the same reduction and thickness. It was observed that hot rolling the samples with 50% or more reduction in thickness were responsible for dynamic recrystallization in the hot-rolled samples and led to grain refinement. Unlike the cold-rolled samples, the hot-rolled samples did not fracture the carbides and appeared to heal the casting defects. The recovery phenomenon was similar to the cold-rolled samples above the eutectoid temperatures, but owing to the refined grain size, the precipitation of the lamellar phase was far more rapid in these samples and the hardness increased more rapidly than in the cold rolled sample when heated below the eutectoid temperature. The data generated from these rolling efforts has been used to make the process modeling efforts more robust and applicable to all USHPRR partner rolling mills. The flow stress for cold rolling the samples was determined to be between 170-190 ksi, with frictional forces between 0.2 and 0.4 for the PNNL mill. The measured roll separation forces and those simulated using finite element methods for hot and cold rolling for the PNNL rolling mill were in good agreement.« less

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.

Materials review for improved automotive gas turbine engine. [superalloys, refractory alloys, and ceramics

NASA Technical Reports Server (NTRS)

Belleau, C.; Ehlers, W. L.; Hagen, F. A.

1978-01-01

The potential role of superalloys, refractory alloys, and ceramics in the hottest sections of engines operating with turbine inlet temperatures as high as 1370 C is examined. The convential superalloys, directionally solidified eutectics, oxide dispersion strenghened alloys, and tungsten fiber reinforced superalloys are reviewed and compared on the basis of maximum turbine blade temperature capability. Improved high temperature protective coatings and special fabrication techniques for these advanced alloys are discussed. Chromium, columbium, molybdenum, tantalum, and tungsten alloys are also reviewed. Molbdenum alloys are found to be the most suitable for mass produced turbine wheels. Various forms and fabrication processes for silicon nitride, silicon carbide, and SIALON's are investigated for use in highstress and medium stress high temperature environments.

CAN HANDLING FIXTURES

DOEpatents

Kelman, Ler.R.; Yaggee, F.L.

1958-08-01

A sleeveless cauning apparatus is described for bonding and canning uranium fuel elements under the surface of a liquid bonding alloy. The can is supported on a pedestal by vertical pegs, and an adjustable collar is placed around the upper, open end of the can, which preferably is flared to assure accurate centering in the fixture and to guide the uranium slug into the can. The fixture with a can in place is then immersed in a liquid aluminum-silicon alloy and the can becomes filled with the liquid alloy. The slug is inserted by a slug guide located vertically above the can opening. The slug settles by gravity into the can, after which a cap is emplaced. A quenching tool lifts the capped can out of the bath by means of a slot provided for it in the pedestal. This apparatus provides a simple means of canning the slug without danger of injury to the uranium metal or the aluminum can.

Neutron absorbing coating for nuclear criticality control

DOEpatents

Mizia, Ronald E.; Wright, Richard N.; Swank, William D.; Lister, Tedd E.; Pinhero, Patrick J.

2007-10-23

A neutron absorbing coating for use on a substrate, and which provides nuclear criticality control is described and which includes a nickel, chromium, molybdenum, and gadolinium alloy having less than about 5% boron, by weight.

CATALYTIC RECOMBINATION OF RADIOLYTIC GASES IN THORIUM OXIDE SLURRIES

DOEpatents

Morse, L.E.

1962-08-01

A method for the coinbination of hydrogen and oxygen in aqueous thorium oxide-uranium oxide slurries is described. A small amount of molybdenum oxide catalyst is provided in the slurry. This catalyst is applicable to the recombination of hydrogen and/or deuterium and oxygen produced by irradiation of the slurries in nuclear reactors. (AEC)

Molybdenum-99 production calculation analysis of SAMOP reactor based on thorium nitrate fuel

NASA Astrophysics Data System (ADS)

Syarip; Togatorop, E.; Yassar

2018-03-01

SAMOP (Subcritical Assembly for Molybdenum-99 Production) has the potential to use thorium as fuel to produce 99Mo after modifying the design, but the production performance has not been discovered yet. A study needs to be done to obtain the correlation between 99Mo production with the mixed fuel composition of uranium and with SAMOP power on the modified SAMOP design. The study aims to obtain the production of 99Mo based thorium nitrate fuel on SAMOP’s modified designs. Monte Carlo N-Particle eXtended (MCNPX) is required to simulate the operation of the assembly by varying the composition of the uranium-thorium nitrate mixed fuel, geometry and power fraction on the SAMOP modified designs. The burnup command on the MCNPX is used to confirm the 99Mo production result. The assembly is simulated to operate for 6 days with subcritical neutron multiplication factor (keff = 0.97-0.99). The neutron multiplication factor of the modified design (keff) is 0.97, the activity obtained from 99Mo is 18.58 Ci at 1 kW power operation.

Complex, Precision Cast Columbium Alloy Gas Turbine Engine Nozzles Coated to Resist Oxidation.

DTIC Science & Technology

1980-04-01

Microstructures of Sprayed Specimens 64 Table 19 NS-4 Coated C129Y Alloy Specimens Weight Bisque Weight Sintered Weight Silicided Weight Pre-Oxidized...choice of another alloy , while perhaps assisting in the foundry process , would not have yielded a mechanical property data base with advantage over...Mo 250 ppm max; Fe 30 ppm max; Al , Ca, C, Si, Cr, Ni, Cu , Mn, Mg and Sn 10 ppm max each). Molybdenum វim powder (02 2000 ppm max; W 250 ppm max; Fe

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)

21 CFR 888.3670 - Shoulder joint metal/polymer/metal nonconstrained or semi-constrained porous-coated uncemented...

Code of Federal Regulations, 2013 CFR

2013-04-01

... as cobalt-chromium-molybdenum (Co-Cr-Mo) and titanium-aluminum-vanadium (Ti-6Al-4V) alloys, and a... Ti-6Al-4V components, beads or fibers of commercially pure titanium or Ti-6Al-4V alloy, or... and Ti-6Al-4V. The humeral component and glenoid backing have a porous coating made of, in the case of...

21 CFR 888.3670 - Shoulder joint metal/polymer/metal nonconstrained or semi-constrained porous-coated uncemented...

Code of Federal Regulations, 2011 CFR

2011-04-01

... as cobalt-chromium-molybdenum (Co-Cr-Mo) and titanium-aluminum-vanadium (Ti-6Al-4V) alloys, and a... Ti-6Al-4V components, beads or fibers of commercially pure titanium or Ti-6Al-4V alloy, or... and Ti-6Al-4V. The humeral component and glenoid backing have a porous coating made of, in the case of...

21 CFR 888.3670 - Shoulder joint metal/polymer/metal nonconstrained or semi-constrained porous-coated uncemented...

Code of Federal Regulations, 2014 CFR

2014-04-01

... as cobalt-chromium-molybdenum (Co-Cr-Mo) and titanium-aluminum-vanadium (Ti-6Al-4V) alloys, and a... Ti-6Al-4V components, beads or fibers of commercially pure titanium or Ti-6Al-4V alloy, or... and Ti-6Al-4V. The humeral component and glenoid backing have a porous coating made of, in the case of...

21 CFR 888.3670 - Shoulder joint metal/polymer/metal nonconstrained or semi-constrained porous-coated uncemented...

Code of Federal Regulations, 2012 CFR

2012-04-01

... as cobalt-chromium-molybdenum (Co-Cr-Mo) and titanium-aluminum-vanadium (Ti-6Al-4V) alloys, and a... Ti-6Al-4V components, beads or fibers of commercially pure titanium or Ti-6Al-4V alloy, or... and Ti-6Al-4V. The humeral component and glenoid backing have a porous coating made of, in the case of...

The Israeli-American International Conference on Applied Metallurgy

DTIC Science & Technology

1976-08-30

conclusions of this study points to the role of molybdenum in improving the corrosion resistance of stainless steels in H 2SO and HCl. These workers feel...density of dislocations and precipitates . Boulger reviewed experience with aluminum and titanium alloys, super-alloys and steels . The demand for higher...available equipment. N. Atzmon and A. Rosen (Technion) applied combined heat treatment and plastic deformation to a maraging (300) steel , and studied

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.

Method for fabricating wrought components for high-temperature gas-cooled reactors and product

DOEpatents

Thompson, Larry D.; Johnson, Jr., William R.

1985-01-01

A method and alloys for fabricating wrought components of a high-temperature gas-cooled reactor are disclosed. These wrought, nickel-based alloys, which exhibit strength and excellent resistance to carburization at elevated temperatures, include aluminum and titanium in amounts and ratios to promote the growth of carburization resistant films while preserving the wrought character of the alloys. These alloys also include substantial amounts of molybdenum and/or tungsten as solid-solution strengtheners. Chromium may be included in concentrations less than 10% to assist in fabrication. Minor amounts of carbon and one or more carbide-forming metals also contribute to high-temperature strength.

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

Rabin, S.A.; Martin, M.M.; Lotts, A.L.

The fabricability of dispersion fuels using UO/sub 2/ or UC as the dispersoid and uranium combined with 10 to 15 wt% Mo as the matrix was investigated. Cores containing l7.8 wt% UO/sub 2/ dispersed in U-- 15 wt.% Mo were successfully fabricated to about 80% of theoretical density by cold pressing at 50 tsi, sintering at 1100 deg C, and cold coining at 50 tsi. Comparable results were obtained with UC as the dispersoid. Core fabrication results varied greatly with the type of matrix powder used. Occluded gases, pour density, and surface cleanliness bore important relations to the fabrication behaviormore » of powders. Suitable pressing and sintering results were obtained with prealloyed, calcium-reduced U--Mo powder and with molybdenum and calcium-reduced uranium as elemental powders. Shotted prealloyed powders were difficult to press and sinter, as were elemental and prealloyed powders prepared by hydriding. The cores containing UO/sub 2/ were picture-frame, hot-roll-clad as miniature plates. Molybdenum, Fansteel 82, and Zr--3 wt% Al were investigated as cladding materials. While each bonded well to itself, only the molybdenum-clad core, rolled at 1150 deg C to 10/1 reduction, resulted in dispersions free of ruptures and UO/sub 2/ fragmentation and in strong bonding to the core, evaluated by metallography, mechanical peel, and thermal shock tests. The matrix phase was homogeneous, but the UO/sub 2/ dispersoid showed stringering characteristic of cores worked by hot rolling. Core densities as high as 99% of theoretical were obtained. (auth)« less

Evaluation of Thermal Stability of Ausferrite in Austempered Ductile Iron Using Differential Scanning Calorimetry

NASA Astrophysics Data System (ADS)

Warsinski, Karl C.

Austempered Ductile Iron (ADI) is prone to changes in microstructure and mechanical properties when exposed to elevated service temperatures. Differential Scanning Calorimetry has been used to evaluate the stabilizing effects of copper, nickel, molybdenum, and cobalt on the ausferrite structure. Previous studies have conflated the effects of various alloy additions, and little effort has been made to systematically catalog the effects of individual elements. The focus of the current research has been to identify alloying elements that more strongly stabilize the ausferrite structure in order to improve service life of ADI at elevated temperatures. Nickel has been shown to have a moderate stabilizing effect, while copper and molybdenum cause a much sharper increase in activation energy. Cobalt has a high stabilizing effect at 0.5% addition by weight, but a further increase to 2.36% results in a slight decrease in activation energy.

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.

Creep of Refractory Fibers and Modeling of Metal and Ceramic Matrix Composite Creep Behavior

NASA Technical Reports Server (NTRS)

Tewari, S.N.

1995-01-01

Our concentration during this research was on the following subprograms. (1) Ultra high vacuum creep tests on 218, ST300 and WHfC tungsten and MoHfC molybdenum alloy wires, temperature range from 1100 K to 1500 K, creep time of 1 to 500 hours. (2) High temperature vacuum tensile tests on 218, ST300 and WHfC tungsten and MoHfC molybdenum alloy wires. (3) Air and vacuum tensile creep tests on polycrystalline and single crystal alumina fibers, such as alumina-mullite Nextel fiber, yttrium aluminum ganet (YAG) and Saphikon, temperature range from 1150 K to 1470 K, creep time of 2 to 200 hours. (4) Microstructural evaluation of crept fibers, TEM study on the crept metal wires, SEM study on the fracture surface of ceramic fibers. (5) Metal Matrix Composite creep models, based on the fiber creep properties and fiber-matrix interface zone formation.

Application of microdynamics and lattice mechanics to problems in plastic flow and fracture. Final report, 1 April 1973--31 March 1978

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

Bilello, J C; Liu, J M

Progress in an investigation of the application of microdynamics and lattice mechanics to the problems in plastic flow and fracture is described. The research program consisted of both theoretical formulations and experimental measurements of a number of intrinsic material parameters in bcc metals and alloys including surface energy, phonon-dispersion curves for dislocated solids, dislocation-point defect interaction energy, slip initiation and microplastic flow behavior. The study has resulted in an improved understanding in the relationship among the experimentally determined fracture surface energy, the intrinsic cohesive energy between atomic planes, and the plastic deformation associated with the initial stages of crack propagation.more » The values of intrinsic surface energy of tungsten, molybdenum, niobium and niobium-molybdenum alloys, deduced from the measurements, serve as a starting point from which fracture toughness of these materials in engineering service may be intelligently discussed.« less

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.

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

DOEpatents

Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

1999-01-01

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

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

Iron-aluminum alloys having high room-temperature and method for making same

DOEpatents

Sikka, V.K.; McKamey, C.G.

1993-08-24

A wrought and annealed iron-aluminum alloy is described consisting essentially of 8 to 9.5% aluminum, an effective amount of chromium sufficient to promote resistance to aqueous corrosion of the alloy, and an alloying constituent selected from the group of elements consisting of an effective amount of molybdenum sufficient to promote solution hardening of the alloy and resistance of the alloy to pitting when exposed to solutions containing chloride, up to about 0.05% carbon with up to about 0.5% of a carbide former which combines with the carbon to form carbides for controlling grain growth at elevated temperatures, and mixtures thereof, and the balance iron, wherein said alloy has a single disordered [alpha] phase crystal structure, is substantially non-susceptible to hydrogen embrittlement, and has a room-temperature ductility of greater than 20%.

Characterization of Molybdate Conversion Coatings for Aluminum Alloys by Electrochemical Impedance Spectroscopy

NASA Technical Reports Server (NTRS)

Calle, Luz Marina

2000-01-01

Electrochemical impedance spectroscopy (EIS) was used to investigate the corrosion inhibiting properties of newly developed proprietary molybdate conversion coatings on aluminum alloy 2024-T3 under immersion in aerated 5% (w/w) NaCl. Corrosion potential and EIS measurements were gathered for six formulations of the coating at several immersion times for two weeks. Nyquist as well as Bode plots of the data were obtained. The conversion-coated alloy panels showed an increase in the corrosion potential during the first 24 hours of immersion that later subsided and approached a steady value. Corrosion potential measurements indicated that formulations A, D, and F exhibit a protective effect on aluminum 2024-T3. The EIS spectra of the conversion-coated alloy were characterized by an impedance that is higher than the impedance of the bare alloy at all the immersion times. The low frequency impedance, Z(sub lf) (determined from the value at 0.05 Hz) for the conversion-coated alloy was higher at all the immersion times than that of the bare panel. This indicates improvement of corrosion resistance with addition of the molybdate conversion coating. Scanning electron microscopy (SEM) revealed the presence of cracks in the coating and the presence of cubic crystals believed to be calcium carbonate. Energy dispersive spectroscopy (EDS) of the test panels revealed the presence of high levels of aluminum, oxygen, and calcium but did not detect the presence of molybdenum on the test panels. X-ray photoelectron spectroscopy (XPS) indicated the presence of less than 0.01 atomic percent molybdenum on the surface of the coating.

Vapor core propulsion reactors

NASA Technical Reports Server (NTRS)

Diaz, Nils J.

1991-01-01

Many research issues were addressed. For example, it became obvious that uranium tetrafluoride (UF4) is a most preferred fuel over uranium hexafluoride (UF6). UF4 has a very attractive vaporization point (1 atm at 1800 K). Materials compatible with UF4 were looked at, like tungsten, molybdenum, rhenium, carbon. It was found that in the molten state, UF4 and uranium attacked most everything, but in the vapor state they are not that bad. Compatible materials were identified for both the liquid and vapor states. A series of analyses were established to determine how the cavity should be designed. A series of experiments were performed to determine the properties of the fluid, including enhancement of the electrical conductivity of the system. CFD's and experimental programs are available that deal with most of the major issues.

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.

The solubility of metals in Pb17Li liquid alloy

NASA Astrophysics Data System (ADS)

Borgstedt, H. U.; Feuerstein, H.

1992-09-01

The solubility data of iron in the eutectic alloy Pb17Li which were evaluated from corrosion tests in a turbulent flow of the molten alloy are discussed in the frame of solubilities of the transition metals in liquid lead. It is shown that the solubility of iron in the alloy is close to that in lead. This is also the fact for several other alloying elements of steels.A comparison of all known data shows that they are in agreement with generally shown trends for the solubility of the transition metals in low melting metals. These trends indicate comparably high solubilities of nickel and manganese in the liquid metals, lower saturation concentrations of vanadium, chromium, iron, and cobalt, and extremely low solubility of molybdenum.

Iron-aluminum alloys having high room-temperature and method for making same

DOEpatents

Sikka, Vinod K.; McKamey, Claudette G.

1993-01-01

Iron-aluminum alloys having selectable room-temperature ductilities of greater than 20%, high resistance to oxidation and sulfidation, resistant pitting and corrosion in aqueous solutions, and possessing relatively high yield and ultimate tensile strengths are described. These alloys comprise 8 to 9.5% aluminum, up to 7% chromium, up to 4% molybdenum, up to 0.05% carbon, up to 0.5% of a carbide former such as zirconium, up to 0.1 yttrium, and the balance iron. These alloys in wrought form are annealed at a selected temperature in the range of 700.degree. C. to about 1100.degree. C. for providing the alloys with selected room-temperature ductilities in the range of 20 to about 29%.

Iron aluminide alloys with improved properties for high temperature applications

DOEpatents

McKamey, Claudette G.; Liu, Chain T.

1990-01-01

An improved iron aluminide alloy of the DO.sub.3 type that has increased room temperature ductility and improved high elevated temperature strength. The alloy system further is resistant to corrosive attack in the environments of advanced energy corrosion systems such as those using fossil fuels. The resultant alloy is relatively inexpensive as contrasted to nickel based and high nickel steels currently utilized for structural components. The alloy system consists essentially of 26-30 at. % aluminum, 0.5-10 at. % chromium, 0.02-0.3 at. % boron plus carbon, up to 2 at. % molybdenum, up to 1 at. % niobium, up to 0.5 at. % zirconium, up to 0.1 at. % yttrium, up to 0.5 at. % vanadium and the balance iron.

Iron aluminide alloys with improved properties for high temperature applications

DOEpatents

McKamey, C.G.; Liu, C.T.

1990-10-09

An improved iron aluminide alloy of the DO[sub 3] type is described that has increased room temperature ductility and improved high elevated temperature strength. The alloy system further is resistant to corrosive attack in the environments of advanced energy conversion systems such as those using fossil fuels. The resultant alloy is relatively inexpensive as contrasted to nickel based and high nickel steels currently utilized for structural components. The alloy system consists essentially of 26--30 at. % aluminum, 0.5--10 at. % chromium, 0.02--0.3 at. % boron plus carbon, up to 2 at. % molybdenum, up to 1 at. % niobium, up to 0.5 at. % zirconium, up to 0.1 at. % yttrium, up to 0.5 at. % vanadium and the balance iron. 3 figs.

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

Molybdenum Availability Is Key to Nitrate Removal in Contaminated Groundwater Environments

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

Thorgersen, Michael P.; Lancaster, W. Andrew; Vaccaro, Brian J.

2015-05-15

The concentrations of molybdenum (Mo) and 25 other metals were measured in groundwater samples from 80 wells on the Oak Ridge Reservation (ORR) (Oak Ridge, TN), many of which are contaminated with nitrate, as well as uranium and various other metals. Moreover, the concentrations of nitrate and uranium were in the ranges of 0.1 μM to 230 mM and <0.2 nM to 580 μM, respectively. Most metals examined had significantly greater median concentrations in a subset of wells that were highly contaminated with uranium (≥126 nM). They included cadmium, manganese, and cobalt, which were 1,300- to 2,700-fold higher. A notablemore » exception, however, was Mo, which had a lower median concentration in the uranium-contaminated wells. This is significant, because Mo is essential in the dissimilatory nitrate reduction branch of the global nitrogen cycle. It is required at the catalytic site of nitrate reductase, the enzyme that reduces nitrate to nitrite. Furthermore, more than 85% of the groundwater samples contained less than 10 nM Mo, whereas concentrations of 10 to 100 nM Mo were required for efficient growth by nitrate reduction for twoPseudomonasstrains isolated from ORR wells and by a model denitrifier,Pseudomonas stutzeriRCH2. Higher concentrations of Mo tended to inhibit the growth of these strains due to the accumulation of toxic concentrations of nitrite, and this effect was exacerbated at high nitrate concentrations. The relevance of these results to a Mo-based nitrate removal strategy and the potential community-driving role that Mo plays in contaminated environments are discussed.« less

Reestablishing Strategic and Critical Material Security in the Department of Defense

DTIC Science & Technology

2011-05-11

Nickel >700% Tungsten 300% Titanium 600% Cobalt 325% Germanium 300% Chromium 500% Molybdenum 500% Indium 300% Manganese 350% Rhenium > 1000% Peak...CHAIN LEADERSHIP New Mission Example • Currently working with Tinker Air Force Base on a rhenium availability issue – Rhenium is a super alloy used in...acquisitions to assure industrial base capability – Titanium – Rare Earth Elements – Germanium – Rhenium / nickel super-alloys – Other materials as supply chain

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.

Chromium modified nickel-iron aluminide useful in sulfur bearing environments

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

Cathcart, J.V.; Liu, C.T.

1989-06-13

This patent describes an improved nickel-iron aluminide containing chromium and molybdenum additions to improve resistance to sulfur attack. The corrosive effects of sulfur are discussed and the chemical composition of corrosion resistant alloys is illustrated.

Evidence for the dissolution of molybdenum during tribocorrosion of CoCrMo hip implants in the presence of serum protein.

PubMed

Simoes, Thiago A; Bryant, Michael G; Brown, Andy P; Milne, Steven J; Ryan, Mary; Neville, Anne; Brydson, Rik

2016-11-01

We have characterized CoCrMo, Metal-on-Metal (MoM) implant, wear debris particles and their dissolution following cycling in a hip simulator, and have related the results to the tribocorrosion of synthetic wear debris produced by milling CoCrMo powders in solutions representative of environments in the human body. Importantly, we have employed a modified ICP-MS sample preparation procedure to measure the release of ions from CoCrMo alloys during wear simulation in different media; this involved use of nano-porous ultrafilters which allowed complete separation of particles from free ions and complexes in solution. As a result, we present a new perspective on the release of metal ions and formation of metal complexes from CoCrMo implants. The new methodology enables the mass balance of ions relative to complexes and particles during tribocorrosion in hip simulators to be determined. A much higher release of molybdenum ions relative to cobalt and chromium has been measured. The molybdenum dissolution was enhanced by the presence of bovine serum albumin (BSA), possibly due to the formation of metal-protein complexes. Overall, we believe that the results could have significant implications for the analysis and interpretation of metal ion levels in fluids extracted from hip arthroplasty patients; we suggest that metal levels, including molybdenum, be analysed in these fluids using the protocol described here. We have developed an important new protocol for the analysis of metal ion levels in fluids extracted from hip implant patients and also hip simulators. Using this procedure, we present a new perspective on the release of metal ions from CoCrMo alloy implants, revealing significantly lower levels of metal ion release during tribocorrosion in hip simulators than previously thought, combined with the release of much higher percentages of molybdenum ions relative to cobalt and chromium. This work is of relevance, both from the perspective of the fundamental science and study of metal-protein interactions, enabling understanding of the ongoing problem associated with the biotribocorrosion and the link to inflammation associated with Metal-on-Metal (MoM) hip implants made from CoCrMo alloys. Copyright © 2016 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

DOEpatents

Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

1999-03-23

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

METHOD OF SUPPRESSING UAl$sub 4$ FORMATION IN U-Al ALLOYS

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1960-08-23

A method is given for suppressing the formation of UAl/sub 4/ in uranium- - aluminum alloys, thereby rendering these alloys more easily workable. The method comprises incorporating in the base alloy a Group Four element selected from the group consisting of Si, Ti, Ge, Zr, and Sn, the addition preferably being within the range of 0.5to20at.%.

U-Zr alloy: XPS and TEM study of surface passivation

NASA Astrophysics Data System (ADS)

Paukov, M.; Tkach, I.; Huber, F.; Gouder, T.; Cieslar, M.; Drozdenko, D.; Minarik, P.; Havela, L.

2018-05-01

Surface reactivity of Uranium metal is an important factor limiting its practical applications. Bcc alloys of U with various transition metals are much less reactive than pure Uranium. So as to specify the mechanism of surface protection, we have been studying the U-20 at.% Zr alloy by photoelectron spectroscopy and transmission electron microscopy. The surface was studied in as-obtained state, in various stages of surface cleaning, and during an isochronal annealing cycle. The analysis based on U-4f, Zr-3p, and O-1 s spectra shows that a Zr-rich phase segregates at the surface at temperatures exceeding 550 K, which provides a self-assembled coating. The comparison of oxygen exposure of the stoichiometric and coated surfaces shows that the coating is efficiently preventing the oxidation of uranium even at elevated temperatures. The coating can be associated with the UZr2+x phase. TEM study indicated that the coating is about 20 nm thick. For the clean state, the U-4f core-level lines of the bcc alloy are practically identical to those of α-U, revealing similar delocalization of the 5f electronic states.

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 to provide additional confidence with the results. The actual corrosion rates of UMo fuel is very likely to be lower than assumed within this report which can be confirmed with additional testing.« 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.

Enhancements to High Temperature In-Pile Thermocouple Performance

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

J. C. Crepeau; J. L. Rempe; J. E. Daw

2008-03-01

A joint University of Idaho (UI) and Idaho National Laboratory (INL) University Nuclear Research Initiative (UNERI) was to initiated to extend initial INL efforts to develop doped lybdenum/niobium alloy High Temperature Irradiation Resistant Thermocouples (HTIR-TCs). The overall objective of this UNERI was to develop recommendations for an optimized thermocouple design for high temperature, long duration, in-pile testing by expanding upon results from initial INL efforts. Tasks to quantify the impact of candidate enhancements, such as alternate alloys, alternate geometries, and alternate thermocouple fabrication techniques, on thermocouple performance were completed at INL's High Temperature Test Laboratory (HTTL), a state of themore » art facility equipped with specialized equipment and trained staff in the area of high temperature instrumentation development and evaluation. Key results of these evaluations, which are documented in this report, are as follows. The doped molybdenum and Nb-1%Zr, which were proposed in the initial INL HTIR-TC design, were found to retain ductility better than the developmental molybdenum-low niobium alloys and the niobium-low molybdenum alloys evaluated. Hence, the performance and lower cost of the commercially available KW-Mo makes a thermocouple containing KW-Mo and Nb-1%Zr the best option at this time. HTIR-TCs containing larger diameter wires offer the potential to increase HTIR-TC stability and reliability at higher temperatures. HTIR-TC heat treatment temperatures and times should be limited to not more than 100 °C above the proposed operating temperatures and to durations of at least 4 to 5 hours. Preliminary investigations suggest that the performance of swaged and loose assembly HTIR-TC designs is similar. However, the swaged designs are less expensive and easier to construct. In addition to optimizing HTIR-TC performance, This UNERI project provided unique opportunities to several University of Idaho students, allowing them to become familiar with the techniques and equipment used for specialized high temperature instrumentation fabrication and evaluation and to author/coauthor several key conference papers and journal articles.« less

Magnesium transport extraction of transuranium elements from LWR fuel

DOEpatents

Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.; Pierce, R. Dean

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.

PROTECTIVELY COVERED ARTICLE AND METHOD OF MANUFACTURE

DOEpatents

Plott, R.F.

1958-10-28

A method of casting a protective jacket about a ura nium fuel element that will bond completely to the uranium without the use of stringers or supports that would ordinarily produce gaps in the cast metal coating and bond is presented. Preformed endcaps of alumlnum alloyed with 13% silicon are placed on the ends of the uranium fuel element. These caps will support the fuel element when placed in a mold. The mold is kept at a ing alloy but below that of uranium so the cast metal jacket will fuse with the endcaps forming a complete covering and bond to the fuel element, which would otherwise oxidize at the gaps or discontinuities lefi in the coating by previous casting methods.

Generation of long time creep data on refractory alloys at elevated temperatures

NASA Technical Reports Server (NTRS)

Sheffler, K. D.

1970-01-01

Creep tests were conducted on two tantalum alloys (ASTAR 811C and T-111 alloy), on a molybdenum alloy (TZM), and on CVD tungsten. The T-111 alloy 1% creep life data have been subjected to Manson's station function analysis, and the progress on this analysis is described. In another test program, the behavior of T-111 alloy with continuously varying temperatures and stresses has been studied. The results indicated that the previously described analysis predicts the observed creep behavior with reasonable accuracy. In addition to the T-111 test program, conventional 1% creep life data have been obtained for ASTAR 811C alloy. Previously observed effects of heat treatment on the creep strength of this material have been discussed and a model involving carbide strengthening primarily at the grain boundaries, rather than in a classical dispersion hardening mechanism, has been proposed to explain the observed results.

Elevated Temperature Tensile Tests on DU–10Mo Rolled Foils

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

Schulthess, Jason

2014-09-01

Tensile mechanical properties for uranium-10 wt.% molybdenum (U–10Mo) foils are required to support modeling and qualification of new monolithic fuel plate designs. It is expected that depleted uranium-10 wt% Mo (DU–10Mo) mechanical behavior is representative of the low enriched U–10Mo to be used in the actual fuel plates, therefore DU-10Mo was studied to simplify material processing, handling, and testing requirements. In this report, tensile testing of DU-10Mo fuel foils prepared using four different thermomechanical processing treatments were conducted to assess the impact of foil fabrication history on resultant tensile properties.

HEAT TREATED U-Mo ALLOY

DOEpatents

McGeary, R.K.; Justusson, W.M.

1960-02-23

A reactor fuel element comprising a gamma-phase alloy consisting of 11 to 16 wt.% of molyhdenum and the balance uranium, annealed between 350 and 525 deg C and quenched to preserve the gamma phase, is reported.

Effects of alloying on aging and hardening processes of steel with 20% nickel

NASA Technical Reports Server (NTRS)

Bogachev, I. N.; Zvigintsev, N. V.; Maslakova, T. M.

1981-01-01

Measurements of hardness, thermal emf, and electrical resistance were used to study the effects of Co, Mo, Ti and Al contents on aging and hardening processes in Fe 20%Ni steel. It is shown that the effects of these alloying elements differ substantially. Anomalies which arise in the temperature dependence of physical properties due to the presence of cobalt and molybdenum are reduced by the inclusion of titanium and aluminum (and vice versa).

Research and Development on Titanium Alloys

DTIC Science & Technology

1949-08-31

present contract was submitted in lieu of the first regular bimonthly progress report. The attached report contains an account of the following: 1 . A...and 5 to 11 per cent, respectively. l. Titanium - 5 per cent molybdenum base alloys with additions of 1 per cent copper, 2 per cent copper, 1 per...cent manganese, and 2 per cent iron, BATTELLE MEMORIAL INSTITUTE TABLE OF CONTENTS SUMMARY. ** * se0* .0. • • 0000 0 C 0, 00 1 INTRODUCTION. . . o

Microstructures in rapidly solidified Ni-Mo alloys

NASA Technical Reports Server (NTRS)

Jayaraman, N.; Tewari, S. N.; Hemker, K. J.; Glasgow, T. K.

1985-01-01

Ni-Mo alloys of compositions ranging from pure Ni to Ni-40 at % Mo were rapidly solidified by Chill Block Melt Spinning in vacuum and were examined by optical metallography, X-ray diffraction and transmission electron microscopy. Rapid solidification resulted in an extension of molybdenum solubility in nickel from 28 to 37.5 at %. A number of different phases and microstructures were seen at different depths (solidification conditions) from the quenched surface of the melt spun ribbons.

Iron rich low cost superalloys. Ph.D. Thesis. Final Report

NASA Technical Reports Server (NTRS)

Wayne, S. F.

1985-01-01

An iron-rich low-cost superalloy was developed. The alloy, when processed by conventional chill casting, has physical and mechanical properties that compare favorably with existing nickel and cobalt based superalloys while containing significantly lower amounts of strategic elements. Studies were also made on the properties of Cr(20)-Mn(10)-C(3.4)-Fe(bal.), a eutectic alloy processed by chill casting and directional solidification which produced an aligned microstructure consisting of M7C3 fibers in a gamma-Fe matrix. Thermal expansion of the M7C3 (M = Fe, Cr, Mn) carbide lattice was measured up to 800 C and found to be highly anisotropic, with the a-axis being the predominant mode of expansion. Repetitive impact sliding wear experiments performed with the Fe rich eutectic alloy showed that the directionally solidified microstructure greatly improved the alloy's wear resistance as compared to the chill cast microstructure and conventional nickel base superalloys. Studies on the molybdenum cementite phase prove that the crystal structure of the xi phase is not orthorhombic. The crystal structure of the xi phase is made up of octahedra building elements consisting of four Mo and two Fe atoms and trigonal prisms consisting of four Fe and two Mo atoms. The voids are occupied by carbon atoms. The previous chemical formula for the molybdenum cementite MoFe2C is now clearly seen to be Mo12Fe22C10.

Hydrogen-assisted post-growth substitution of tellurium into molybdenum disulfide monolayers with tunable compositions

NASA Astrophysics Data System (ADS)

Yin, Guoli; Zhu, Dancheng; Lv, Danhui; Hashemi, Arsalan; Fei, Zhen; Lin, Fang; Krasheninnikov, Arkady V.; Zhang, Ze; Komsa, Hannu-Pekka; Jin, Chuanhong

2018-04-01

Herein we report the successful doping of tellurium (Te) into molybdenum disulfide (MoS2) monolayers to form MoS2x Te2(1-x) alloy with variable compositions via a hydrogen-assisted post-growth chemical vapor deposition process. It is confirmed that H2 plays an indispensable role in the Te substitution into as-grown MoS2 monolayers. Atomic-resolution transmission electron microscopy allows us to determine the lattice sites and the concentration of introduced Te atoms. At a relatively low concentration, tellurium is only substituted in the sulfur sublattice to form monolayer MoS2(1-x)Te2x alloy, while with increasing Te concentration (up to ˜27.6% achieved in this study), local regions with enriched tellurium, large structural distortions, and obvious sulfur deficiency are observed. Statistical analysis of the Te distribution indicates the random substitution. Density functional theory calculations are used to investigate the stability of the alloy structures and their electronic properties. Comparison with experimental results indicate that the samples are unstrained and the Te atoms are predominantly substituted in the top S sublattice. Importantly, such ultimately thin Janus structure of MoS2(1-x)Te2x exhibits properties that are distinct from their constituents. We believe our results will inspire further exploration of the versatile properties of asymmetric 2D TMD alloys.

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)

DUCTILE URANIUM FUEL FOR NUCLEAR REACTORS AND METHOD OF MAKING

DOEpatents

Zegler, S.T.

1963-11-01

The fabrication process for a ductile nuclear fuel alloy consisting of uranium, fissium, and from 0.25 to 1.0 wt% of silicon or aluminum or from 0.25 to 2 wt% of titanium or yttrium is presented. (AEC)

The Effects of Alloy Chemistry on Localized Corrosion of Austenitic Stainless Steels

NASA Astrophysics Data System (ADS)

Sapiro, David O.

This study investigated localized corrosion behavior of austenitic stainless steels under stressed and unstressed conditions, as well as corrosion of metallic thin films. While austenitic stainless steels are widely used in corrosive environments, they are vulnerable to pitting and stress corrosion cracking (SCC), particularly in chloride-containing environments. The corrosion resistance of austenitic stainless steels is closely tied to the alloying elements chromium, nickel, and molybdenum. Polarization curves were measured for five commercially available austenitic stainless steels of varying chromium, nickel, and molybdenum content in 3.5 wt.% and 25 wt.% NaCl solutions. The alloys were also tested in tension at slow strain rates in air and in a chloride environment under different polarization conditions to explore the relationship between the extent of pitting corrosion and SCC over a range of alloy content and environment. The influence of alloy composition on corrosion resistance was found to be consistent with the pitting resistance equivalent number (PREN) under some conditions, but there were also conditions under which the model did not hold for certain commercial alloy compositions. Monotonic loading was used to generate SCC in in 300 series stainless steels, and it was possible to control the failure mode through adjusting environmental and polarization conditions. Metallic thin film systems of thickness 10-200 nm are being investigated for use as corrosion sensors and protective coatings, however the corrosion properties of ferrous thin films have not been widely studied. The effects of film thickness and substrate conductivity were examined using potentiodynamic polarization and scanning vibrating electrode technique (SVET) on iron thin films. Thicker films undergo more corrosion than thinner films in the same environment, though the corrosion mechanism is the same. Conductive substrates encourage general corrosion, similar to that of bulk iron, while insulating substrates supported only localized corrosion.

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

METAL COATING BATHS

DOEpatents

Robinson, J.W.

1958-08-26

A method is presented for restoring the effectiveness of bronze coating baths used for hot dip coating of uranium. Such baths, containing a high proportion of copper, lose their ability to wet uranium surfaces after a period of use. The ability of such a bath to wet uranium can be restored by adding a small amount of metallic aluminum to the bath, and skimming the resultant hard alloy from the surface.

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.

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.

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

Study of Ni-Mo electrodeposition in direct and pulse-reverse current

NASA Astrophysics Data System (ADS)

Stryuchkova, Yu M.; Rybin, N. B.; Suvorov, D. V.; Gololobov, G. P.; Tolstoguzov, A. B.; Tarabrin, D. Yu; Serpova, M. A.; Korotchenko, V. A.; Slivkin, E. V.

2017-05-01

Process of electrochemical deposition of the coating based on a binary nickel-molybdenum alloy onto a nickel substrate under pulse mode with current reverse within the range of current density change from 2 to 9 A/dm2 has been researched. Coating structure and its surface morphology have been studied. Method of X-ray energy dispersive spectroscopy has determined a percentage ratio of alloy components in the coating. Mode to obtain the densest and smoothest deposits has been identified under considered terms.

Mechanical, Corrosion, and Fatigue Properties of 15-5 PH, Inconel 718, and Rene 41 Weldments

DTIC Science & Technology

1975-05-01

TUCUMvARI (PGH 2) It was originally developed for high - temperature use by Armco Steel Corporation. 2 High strength is obtained by the precipitation...Rene 41 was developed by General Electric Company as a high - temperature turbine alloy. It is a nickel-base alloy, high in chromium, cobalt, and... molybdenum . Its high strength comes from the precipitation of a gamma-prime phase consisting of Ni3AI and Ni5Ti, and from the solid solution effects of

Investigation of High Temperature Ductility Losses in Alpha-Beta Titanium Alloys

DTIC Science & Technology

1988-04-01

Gleeble simulation of GTAW thermal _ cycles, Figure 1.1 (6). They found that Ti-6AI-4V (Ti-64), Ti-6A1-2Nb-lTa-0.8Mo (Ti-6211), and Ti-6AI suffered...or weak beta stabilizers depending on the other alloying elements present. Vanadium, molybdenum, tantalum, niobium, chromium , silicon, copper...elements. Chromium , - silicon, copper, manganese, cobalt, iron, and hydrogen are all eutectic formers. A schematic binary phase diagram of a 0 beta

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.

Study of fluoride corrosion of nickel alloys

NASA Technical Reports Server (NTRS)

Gunther, W. H.; Steindler, M. J.

1969-01-01

Report contains the results of an investigation of the corrosion resistance of nickel and nickel alloys exposed to fluorine, uranium hexafluoride, and volatile fission product fluorides at high temperatures. Survey of the unclassified literature on the subject is included.

Method of preparing copper-dendritic composite alloys for mechanical reduction

DOEpatents

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

1988-01-01

Copper-dendritic composite alloys are prepared for mechanical reduction to increase tensile strength by dispersing molten droplets of the composite alloy into an inert gas; solidifying the droplets in the form of minute spheres or platelets; and compacting a mass of the spheres or platelets into an integrated body. The spheres preferably have diameters of from 50 to 2000 .mu.m, and the platelets thicknesses of 100 to 2000 .mu.m. The resulting spheres or platelets will contain ultra-fine dendrites which produce higher strengths on mechanical reduction of the bodies formed therefrom, or comparable strengths at lower reduction values. The method is applicable to alloys of copper with vanadium, niobium, tantalum, chromium, molybdenum, tungsten, iron and cobalt.

Method of preparing copper-dendritic composite alloys for mechanical reduction

DOEpatents

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

1988-09-13

Copper-dendritic composite alloys are prepared for mechanical reduction to increase tensile strength by dispersing molten droplets of the composite alloy into an inert gas; solidifying the droplets in the form of minute spheres or platelets; and compacting a mass of the spheres or platelets into an integrated body. The spheres preferably have diameters of from 50 to 2,000 [mu]m, and the platelets thicknesses of 100 to 2,000 [mu]m. The resulting spheres or platelets will contain ultra-fine dendrites which produce higher strengths on mechanical reduction of the bodies formed therefrom, or comparable strengths at lower reduction values. The method is applicable to alloys of copper with vanadium, niobium, tantalum, chromium, molybdenum, tungsten, iron and cobalt. 3 figs.

Radionuclide deposition control

DOEpatents

Brehm, William F.; McGuire, Joseph C.

1980-01-01

The deposition of radionuclides manganese-54, cobalt-58 and cobalt-60 from liquid sodium coolant is controlled by providing surfaces of nickel or high nickel alloys to extract the radionuclides from the liquid sodium, and by providing surfaces of tungsten, molybdenum or tantalum to prevent or retard radionuclide deposition.

Characterization of Brazed Joints of C-C Composite to Cu-clad-Molybdenum

NASA Technical Reports Server (NTRS)

Singh, M.; Asthana, R.

2008-01-01

Carbon-carbon composites with either pitch+CVI matrix or resin-derived matrix were joined to copper-clad molybdenum using two active braze alloys, Cusil-ABA (1.75% Ti) and Ticusil (4.5% Ti). The brazed joints revealed good interfacial bonding, preferential precipitation of Ti at the composite/braze interface, and a tendency toward de-lamination in resin-derived C-C composite due to its low inter-laminar shear strength. Extensive braze penetration of the inter-fiber channels in the pitch+CVI C-C composites was observed. The relatively low brazing temperatures (<950 C) precluded melting of the clad layer and restricted the redistribution of alloying elements but led to metallurgically sound composite joints. The Knoop microhardness (HK) distribution across the joint interfaces revealed sharp gradients at the Cu-clad-Mo/braze interface and higher hardness in Ticusil (approx.85-250 HK) than in Cusil-ABA (approx.50-150 HK). These C-C/Cu-clad-Mo joints with relatively low thermal resistance may be promising for thermal management applications.

Relative toxicity of lead and selected substitute shot types to game farm mallards

USGS Publications Warehouse

Irby, H.D.; Locke, L.N.; Bagley, George E.

1967-01-01

The acute toxicity of lead, three types of plastic-coated lead, two lead-magnesium alloys, iron, copper, zinc-coated iron, and molybdenum-coated iron shot were tested in year-old male game farm mallards. Mallards (Anus platyrhynchos) were fed eight number 6 shot of each type and observed for a period of 60 days. Ducks used totaled 230 and most shot types were tested in three replicates of 8 ducks each. Mortality and losses of body weight were the criteria used for judging toxicity. Three types of plastic-coated lead shot were as toxic (93 percent) as the commercial lead shot (96 percent). The average mortality in mallards fed lead-magnesium alloy shot was less (58 percent) than that occurring in birds fed commercial lead shot. Mortality among mallards fed iron, copper, zinc-coated iron or molybdenum-coated iron shot was significantly less than in birds fed lead shot, and was not significantly greater than the conrtols.

Process for making a martensitic steel alloy fuel cladding product

DOEpatents

Johnson, Gerald D.; Lobsinger, Ralph J.; Hamilton, Margaret L.; Gelles, David S.

1990-01-01

This is a very narrowly defined martensitic steel alloy fuel cladding material for liquid metal cooled reactors, and a process for making such a martensitic steel alloy material. The alloy contains about 10.6 wt. % chromium, about 1.5 wt. % molybdenum, about 0.85 wt. % manganese, about 0.2 wt. % niobium, about 0.37 wt. % silicon, about 0.2 wt. % carbon, about 0.2 wt. % vanadium, 0.05 maximum wt. % nickel, about 0.015 wt. % nitrogen, about 0.015 wt. % sulfur, about 0.05 wt. % copper, about 0.007 wt. % boron, about 0.007 wt. % phosphorous, and with the remainder being essentially iron. The process utilizes preparing such an alloy and homogenizing said alloy at about 1000.degree. C. for 16 hours; annealing said homogenized alloy at 1150.degree. C. for 15 minutes; and tempering said annealed alloy at 700.degree. C. for 2 hours. The material exhibits good high temperature strength (especially long stress rupture life) at elevated temperature (500.degree.-760.degree. C.).

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.

Coated Metal Articles and Method of Making

DOEpatents

Boller, Ernest R.; Eubank, Lowell D.

2004-07-06

The method of protectively coating metallic uranium which comprises dipping the metallic uranium in a molten alloy comprising about 20-75% of copper and about 80-25% of tin, dipping the coated uranium promptly into molten tin, withdrawing it from the molten tin and removing excess molten metal, thereupon dipping it into a molten metal bath comprising aluminum until it is coated with this metal, then promptly withdrawing it from the bath.

FUEL ELEMENTS AND METHOD OF MAKING

DOEpatents

Noland, R.A.; Marzano, C.

1958-08-19

A process is described of surface-impregnating bodies of metallic uranium with silicon. Silicon metal is added to or admixed with alkali metal selected from the group consisting of sodiunn, potassium, and sodiunnpotassium alloy. The uraniunn body is then immersed in the mixture obtained and the temperature is raised to between 425 and 600 deg C. The silicon is dissolved and deposits as a uranium-silicon compound on the uranium body.

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

DOE PAGES

Collette, R.; King, J.; Buesch, C.; ...

2016-04-01

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

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

Collette, R.; King, J.; Buesch, C.

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Application of the DART Code for the Assessment of Advanced Fuel Behavior

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

Rest, J.; Totev, T.

2007-07-01

The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO{sub 2}more » fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)« less

48 CFR 252.225-7008 - Restriction on Acquisition of Specialty Metals.

Code of Federal Regulations, 2013 CFR

2013-10-01

... consolidation of non-melt derived metal powders. Specialty metal means— (i) Steel— (A) With a maximum alloy..., chromium, cobalt, molybdenum, nickel, niobium (columbium), titanium, tungsten, or vanadium; (ii) Metal... of Specialty Metals. 252.225-7008 Section 252.225-7008 Federal Acquisition Regulations System DEFENSE...

48 CFR 252.225-7008 - Restriction on Acquisition of Specialty Metals.

Code of Federal Regulations, 2014 CFR

2014-10-01

... consolidation of non-melt derived metal powders. Specialty metal means— (i) Steel— (A) With a maximum alloy..., chromium, cobalt, molybdenum, nickel, niobium (columbium), titanium, tungsten, or vanadium; (ii) Metal... of Specialty Metals. 252.225-7008 Section 252.225-7008 Federal Acquisition Regulations System DEFENSE...

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

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.

DIMENSIONALLY STABLE, CORROSION RESISTANT NUCLEAR FUEL

DOEpatents

Kittel, J.H.

1963-10-31

A method of making a uranium alloy of improved corrosion resistance and dimensional stability is described. The alloy contains from 0-9 weight per cent of an additive of zirconium and niobium in the proportions by weight of 5 to 1 1/ 2. The alloy is cold rolled, heated to two different temperatures, air-cooled, heated to a third temperature, and quenched in water. (AEC)

NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

DOEpatents

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

1957-11-12

This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL

DOEpatents

Buyers, A.G.

1959-06-30

A pyro-chemical method is presented for decontaminating neutron irradiated uranium and separating plutonium therefrom by contact in the molten state with a metal chloride salt. Uranium trichloride and uranium tetrachloride either alone or in admixture with alkaline metal and alkaline eanth metal fluorides under specified temperature and specified phase ratio conditions extract substantially all of the uranium from the irradiated uranium fuel together with certain fission products. The phases are then separated leaving purified uranium metal. The uranium and plutonium in the salt phase can be reduced to forin a highly decontaminated uraniumplutonium alloy. The present method possesses advantages for economically decontaminating irradiated nuclear fuel elements since irradiated fuel may be proccessed immediately after withdrawal from the reactor and the uranium need not be dissolved and later reduced to the metallic form. Accordingly, the uranium may be economically refabricated and reinserted into the reactor.

Ceramic-metal composite article and joining method

DOEpatents

Kang, Shinhoo; Selverian, John H.; Kim, Hans J.; Dunn, Edmund M.; Kim, Kyung S.

1992-01-01

A ceramic-metal article including a ceramic rod, a metal rod, and a braze joining the ceramic and metal rods at a braze area of a coaxial bore in the metal rod. The bore gradually decreases in diameter, having an inward seat area sized for close sliding fit about the ceramic, a larger brazing area near the joint end, and a void area intermediate the braze and seat areas. The ceramic is seated without brazing in the bore seat area. The side wall between the brazing area and the metal outer surface is about 0.030-0.080 inch. The braze includes an inner braze layer, an outer braze layer, and an interlayer about 0.030-0.090 inch thick. A shoulder between the brazing and void areas supports the interlayer during bonding while preventing bonding between the void area and the ceramic member, leaving a void space between the void area and the ceramic member. A venting orifice extends generally radially through the metal member from the outer surface to the void space. The braze layers are palladium, platinum, gold, silver, copper, nickel, indium, chromium, molybdenum, niobium, iron, aluminum, or alloys thereof. Preferred is a gold-palladium-nickel brazing alloy. The interlayer is nickel, molybdenum, copper, tantalum, tungsten, niobium, aluminum, cobalt, iron, or an alloy thereof.

Ceramic-metal composite article and joining method

DOEpatents

Kang, S.; Selverian, J.H.; Kim, H.J.; Dunn, E.M.; Kim, K.S.

1992-04-28

A ceramic-metal article including a ceramic rod, a metal rod, and a braze joining the ceramic and metal rods at a braze area of a coaxial bore in the metal rod is described. The bore gradually decreases in diameter, having an inward seat area sized for close sliding fit about the ceramic, a larger brazing area near the joint end, and a void area intermediate the braze and seat areas. The ceramic is seated without brazing in the bore seat area. The side wall between the brazing area and the metal outer surface is about 0.030-0.080 inch. The braze includes an inner braze layer, an outer braze layer, and an interlayer about 0.030-0.090 inch thick. A shoulder between the brazing and void areas supports the interlayer during bonding while preventing bonding between the void area and the ceramic member, leaving a void space between the void area and the ceramic member. A venting orifice extends generally radially through the metal member from the outer surface to the void space. The braze layers are palladium, platinum, gold, silver, copper, nickel, indium, chromium, molybdenum, niobium, iron, aluminum, or alloys thereof. Preferred is a gold-palladium-nickel brazing alloy. The interlayer is nickel, molybdenum, copper, tantalum, tungsten, niobium, aluminum, cobalt, iron, or an alloy thereof. 4 figs.

Design of high-strength refractory complex solid-solution alloys

DOE PAGES

Singh, Prashant; Sharma, Aayush; Smirnov, A. V.; ...

2018-03-28

Nickel-based superalloys and near-equiatomic high-entropy alloys containing molybdenum are known for higher temperature strength and corrosion resistance. Yet, complex solid-solution alloys offer a huge design space to tune for optimal properties at slightly reduced entropy. For refractory Mo-W-Ta-Ti-Zr, we showcase KKR electronic structure methods via the coherent-potential approximation to identify alloys over five-dimensional design space with improved mechanical properties and necessary global (formation enthalpy) and local (short-range order) stability. Deformation is modeled with classical molecular dynamic simulations, validated from our first-principle data. We predict complex solid-solution alloys of improved stability with greatly enhanced modulus of elasticity (3× at 300 K)more » over near-equiatomic cases, as validated experimentally, and with higher moduli above 500 K over commercial alloys (2.3× at 2000 K). We also show that optimal complex solid-solution alloys are not described well by classical potentials due to critical electronic effects.« less

Microstructural characterization of low and high carbon CoCrMo alloy nanoparticles produced by mechanical milling

NASA Astrophysics Data System (ADS)

Simoes, T. A.; Goode, A. E.; Porter, A. E.; Ryan, M. P.; Milne, S. J.; Brown, A. P.; Brydson, R. M. D.

2014-06-01

CoCrMo alloys are utilised as the main material in hip prostheses. The link between this type of hip prosthesis and chronic pain remains unclear. Studies suggest that wear debris generated in-vivo may be related to post-operative complications such as inflammation. These alloys can contain different amounts of carbon, which improves the mechanical properties of the alloy. However, the formation of carbides could become sites that initiate corrosion, releasing ions and/or particles into the human body. This study analysed the mechanical milling of alloys containing both high and low carbon levels in relevant biological media, as an alternative route to generate wear debris. The results show that low carbon alloys produce significantly more nanoparticles than high carbon alloys. During the milling process, strain induces an fcc to hcp phase transformation. Evidence for cobalt and molybdenum dissolution in the presence of serum was confirmed by ICP-MS and TEM EDX techniques.

Design of high-strength refractory complex solid-solution alloys

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

Singh, Prashant; Sharma, Aayush; Smirnov, A. V.

Nickel-based superalloys and near-equiatomic high-entropy alloys containing molybdenum are known for higher temperature strength and corrosion resistance. Yet, complex solid-solution alloys offer a huge design space to tune for optimal properties at slightly reduced entropy. For refractory Mo-W-Ta-Ti-Zr, we showcase KKR electronic structure methods via the coherent-potential approximation to identify alloys over five-dimensional design space with improved mechanical properties and necessary global (formation enthalpy) and local (short-range order) stability. Deformation is modeled with classical molecular dynamic simulations, validated from our first-principle data. We predict complex solid-solution alloys of improved stability with greatly enhanced modulus of elasticity (3× at 300 K)more » over near-equiatomic cases, as validated experimentally, and with higher moduli above 500 K over commercial alloys (2.3× at 2000 K). We also show that optimal complex solid-solution alloys are not described well by classical potentials due to critical electronic effects.« less

The Gas Hills uranium district and some probable controls for ore deposition

USGS Publications Warehouse

Zeller, Howard Davis

1957-01-01

Uranium deposits occur in the upper coarse-grained facies of the Wind River formation of Eocene age in the Gas Hills district of the southern part of the Wind River Basin. Some of the principal deposits lie below the water table in the unoxidized zone and consist of uraninite and coffinite occurring as interstitial fillings in irregular blanket-like bodies. In the near-surface deposits that lie above the water table, the common yellow uranium minerals consist of uranium phosphates, silicates, and hydrous oxides. The black unoxidized uraninite -coffinite ores show enrichment of molybdenum, arsenic, and selenium when compared to the barren sandstone. Probable geologic controls for ore deposits include: 1) permeable sediments that allowed passage of ore-bearing solutions; 2) numerous faults that acted as impermeable barriers impounding the ore -bearing solutions; 3) locally abundant pyrite, carbonaceous material, and natuial gas containing hydrogen sulfide that might provide a favorable environment for precipitation of uranium. Field and laboratory evidence indicate that the uranium deposits in the Gas Hills district are very young and related to the post-Miocene to Pleistocene regional tilting to the south associated with the collapse of the Granite Mountains fault block. This may have stopped or reversed ground water movement from a northward (basinward) direction and alkaline ground water rich in carbonate could have carried the uranium into the favorable environment that induced precipitation.

Corrosion Characterization in Nickel Plated 110 ksi Low Alloy Steel and Incoloy 925: An Experimental Case Study

NASA Astrophysics Data System (ADS)

Thomas, Kiran; Vincent, S.; Barbadikar, Dipika; Kumar, Shresh; Anwar, Rebin; Fernandes, Nevil

2018-04-01

Incoloy 925 is an age hardenable Nickel-Iron-Chromium alloy with the addition of Molybdenum, Copper, Titanium and Aluminium used in many applications in oil and gas industry. Nickel alloys are preferred mostly in corrosive environments where there is high concentration of H2S, CO2, chlorides and free Sulphur as sufficient nickel content provides protection against chloride-ion stress-corrosion cracking. But unfortunately, Nickel alloys are very expensive. Plating an alloy steel part with nickel would cost much lesser than a part make of nickel alloy for large quantities. A brief study will be carried out to compare the performance of nickel plated alloy steel with that of an Incoloy 925 part by conducting corrosion tests. Tests will be carried out using different coating thicknesses of Nickel on low alloy steel in 0.1 M NaCl solution and results will be verified. From the test results we can confirm that Nickel plated low alloy steel is found to exhibit fairly good corrosion in comparison with Incoloy 925 and thus can be an excellent candidate to replace Incoloy materials.

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

2017-01-01

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the Nevada National Security Site. Fabrication, assembly, and testing of the depleted uranium core has allowed for higher fidelity system level testing at GRC, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

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.

The Electrochemical Behavior of Mo-Ta Alloy in Phosphoric Acid Solution for TFT-LCD Application.

PubMed

Lee, Sang-Hyuk; Kim, Byoung O; Seo, Jong Hyun

2015-10-01

Molybdenum-tantalum alloy thin film is a suitable material for the higher corrosion resistance and low resistivity for gate and data metal lines. In this study, Mo-Ta alloy thin films were prepared by using a DC magnetron co-sputtering system on a glass substrate. An abrupt increase in the etching rates of low Mo-Ta alloys was observed. From the observed impedance analysis, the defect densities in the MoTa oxide films increased from 5.4 x 10(21) (cm(-3)) to 8.02 x 10(21) (cm(-3)) up to the 6 at% of tantalum level; and above the 6 at% of tantalum level, the defect densities decreased. This electrochemical behavior is explained by the mechanical instability of the MoTa oxide film.

Pumped lithium loop test to evaluate advanced refractory metal alloys and simulated nuclear fuel elements

NASA Technical Reports Server (NTRS)

Brandenburf, G. P.; Hoffman, E. E.; Smith, J. P.

1974-01-01

The performance was determined of refractory metal alloys and uranium nitride fuel element specimens in flowing 1900F (1083C) lithium. The results demonstrate the suitability of the selected materials to perform satisfactorily from a chemical compatibility standpoint.

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 within safety limits; however, for 7.9542 g and 8.838 g (× 106 cm-1) the limits were exceeded.

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.

Investigation of molybdate melts as an alternative method of reprocessing used nuclear fuel

DOE PAGES

Hames, Amber L.; Tkac, Peter; Paulenova, Alena; ...

2017-01-17

Here, an investigation of molybdate melts containing sodium molybdate (Na 2MoO 4) and molybdenum trioxide (MoO 3) to achieve the separation of uranium from fission products by crystallization has been performed. The separation is based on the difference in solubility of the fission product metal oxides compared to the uranium oxide or molybdate in the molybdate melt. The molybdate melt dissolves uranium dioxide at high temperatures, and upon cooling, uranium precipitates as uranium dioxide or molybdate, whereas the fission product metals remain soluble in the melt. Small-scale experiments using gram quantities of uranium dioxide have been performed to investigate themore » feasibility of UO 2 purification from the fission products. The composition of the uranium precipitate as well as data for partitioning of several fission product surrogates between the uranium precipitate and molybdate melt for various melt compositions are presented and discussed. The fission products Cs, Sr, Ru and Rh all displayed very large distribution ratios. The fission products Zr, Pd, and the lanthanides also displayed good distribution ratios (D > 10). A melt consisting of 20 wt% MoO 3-50 wt% Na 2MoO 4-30 wt% UO 2 heated to 1313 K and cooled to 1123 K for the physical separation of the UO 2 product from the melt, and washed once with Na 2MoO 4 displays optimum conditions for separation of the UO 2 from the fission products.« 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.

FUEL COMPOSITION FOR NUCLEAR REACTORS

DOEpatents

Andersen, J.C.

1963-08-01

A process for making refractory nuclear fuel elements involves heating uranium and silicon powders in an inert atmosphere to 1600 to 1800 deg C to form USi/sub 3/; adding silicon carbide, carbon, 15% by weight of nickel and aluminum, and possibly also molybdenum and silicon powders; shaping the mixture; and heating to 1700 to 2050 deg C again in an inert atmosphere. Information on obtaining specific compositions is included. (AEC)

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.

Alloys compatibility in molten salt fluorides: Kurchatov Institute related experience

NASA Astrophysics Data System (ADS)

Ignatiev, Victor; Surenkov, Alexandr

2013-10-01

In the last several years, there has been an increased interest in the use of high-temperature molten salt fluorides in nuclear power systems. For all molten salt reactor designs, materials selection is a very important issue. This paper summarizes results, which led to selection of materials for molten salt reactors in Russia. Operating experience with corrosion thermal convection loops has demonstrated good capability of the “nickel-molybdenum alloys + fluoride salt fueled by UF4 and PuF3 + cover gas” system up to 750 °C. A brief description is given of the container material work in progress. Tellurium corrosion of Ni-based alloys in stressed and unloaded conditions studies was also tested in different molten salt mixtures at temperatures up to 700-750 °C, also with measurement of the redox potential. HN80MTY alloy with 1% added Al is the most resistant to tellurium intergranular cracking of Ni-base alloys under study.

Highly Enriched Uranium Metal Cylinders Surrounded by Various Reflector Materials

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

Bernard Jones; J. Blair Briggs; Leland Monteirth

A series of experiments was performed at Los Alamos Scientific Laboratory in 1958 to determine critical masses of cylinders of Oralloy (Oy) reflected by a number of materials. The experiments were all performed on the Comet Universal Critical Assembly Machine, and consisted of discs of highly enriched uranium (93.3 wt.% 235U) reflected by half-inch and one-inch-thick cylindrical shells of various reflector materials. The experiments were performed by members of Group N-2, particularly K. W. Gallup, G. E. Hansen, H. C. Paxton, and R. H. White. This experiment was intended to ascertain critical masses for criticality safety purposes, as well asmore » to compare neutron transport cross sections to those obtained from danger coefficient measurements with the Topsy Oralloy-Tuballoy reflected and Godiva unreflected critical assemblies. The reflector materials examined in this series of experiments are as follows: magnesium, titanium, aluminum, graphite, mild steel, nickel, copper, cobalt, molybdenum, natural uranium, tungsten, beryllium, aluminum oxide, molybdenum carbide, and polythene (polyethylene). Also included are two special configurations of composite beryllium and iron reflectors. Analyses were performed in which uncertainty associated with six different parameters was evaluated; namely, extrapolation to the uranium critical mass, uranium density, 235U enrichment, reflector density, reflector thickness, and reflector impurities. In addition to the idealizations made by the experimenters (removal of the platen and diaphragm), two simplifications were also made to the benchmark models that resulted in a small bias and additional uncertainty. First of all, since impurities in core and reflector materials are only estimated, they are not included in the benchmark models. Secondly, the room, support structure, and other possible surrounding equipment were not included in the model. Bias values that result from these two simplifications were determined and associated uncertainty in the bias values were included in the overall uncertainty in benchmark keff values. Bias values were very small, ranging from 0.0004 ?k low to 0.0007 ?k low. Overall uncertainties range from ? 0.0018 to ? 0.0030. Major contributors to the overall uncertainty include uncertainty in the extrapolation to the uranium critical mass and the uranium density. Results are summarized in Figure 1. Figure 1. Experimental, Benchmark-Model, and MCNP/KENO Calculated Results The 32 configurations described and evaluated under ICSBEP Identifier HEU-MET-FAST-084 are judged to be acceptable for use as criticality safety benchmark experiments and should be valuable integral benchmarks for nuclear data testing of the various reflector materials. Details of the benchmark models, uncertainty analyses, and final results are given in this paper.« less

High density dispersion fuel

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

Hofman, G.L.

1996-09-01

A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm{sup 3} of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm{sup {minus}3} with U{sub 3}Si{sub 2} as fuel. High-density uranium compounds offer no real density advantage over U{sub 3}Si{sub 2} and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U{sub 3}Si has approximately a 30% higher uranium density but the density of the U{sub 6}X compounds would yield the factormore » 1.5 needed to achieve 9 g cm{sup {minus}3} uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure {alpha}-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic {gamma} phase at low temperatures where normally {alpha} phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing.« less

Cubic γ-phase U-Mo alloys synthesized by splat-cooling

NASA Astrophysics Data System (ADS)

Kim-Ngan, Nhu-T. H.; Tkach, I.; Mašková, S.; Havela, L.; Warren, A.; Scott, T.

2013-09-01

U-Mo alloys are the most promising materials fulfilling the requirements of using low enriched uranium (LEU) fuel in research reactors. From a fundamental standpoint, it is of interest to determine the basic thermodynamic properties of the cubic γ-phase U-Mo alloys. We focus our attention on the use of Mo doping together with ultrafast cooling (with high cooling rates ⩾106 K s-1), which helps to maintain the cubic γ-phase in U-Mo system to low temperatures and on determination of the low-temperature properties of these γ-U alloys. Using a splat cooling method it has been possible to maintain some fraction of the high-temperature γ-phase at room temperature in pure uranium. U-13 at.% Mo splat clearly exhibits the pure γ-phase structure. All the splats become superconducting with Tc in the range from 1.24 K (pure U splat) to 2.11 K (U-15 at.% Mo). The γ-phase in U-Mo alloys undergoes eutectoid decomposition to form equilibrium phases of orthorhombic α-uranium and tetragonal γ‧-phase upon annealing at 500 °C, while annealing at 800 °C has stabilized the initial γ phase. The α-U easily absorbs a large amount of hydrogen (UH3 hydride), while the cubic bcc phase does not absorb any detectable amount of hydrogen at pressures below 1 bar and at room temperature. At 80 bar, the U-15 at.% Mo splat becomes powder consisting of elongated particles of 1-2 mm, revealing amorphous state.

Improving tribological properties of Ti-5Zr-3Sn-5Mo-15Nb alloy by double glow plasma surface alloying

NASA Astrophysics Data System (ADS)

Guo, Lili; Qin, Lin; Kong, Fanyou; Yi, Hong; Tang, Bin

2016-12-01

Molybdenum, an alloying element, was deposited and diffused on Ti-5Zr-3Sn-5Mo-15Nb (TLM) substrate by double glow plasma surface alloying technology at 900, 950 and 1000 °C. The microstructure, composition distribution and micro-hardness of the Mo modified layers were analyzed. Contact angles on deionized water and wear behaviors of the samples against corundum balls in simulated human body fluids were investigated. Results show that the surface microhardness is significantly enhanced after alloying and increases with treated temperature rising, and the contact angles are lowered to some extent. More importantly, compared to as-received TLM alloy, the Mo modified samples, especially the one treated at 1000 °C, exhibit the significant improvement of tribological properties in reciprocating wear tests, with lower specific wear rate and friction coefficient. To conclude, Mo alloying treatment is an effective approach to obtain excellent comprehensive properties including optimal wear resistance and improved wettability, which ensure the lasting and safety application for titanium alloys as the biomedical implants.

O-Pu-U (Oxygen-Plutonium-Uranium)

NASA Astrophysics Data System (ADS)

Materials Science International Team MSIT

This document is part of Subvolume C4 'Non-Ferrous Metal Systems. Part 4: Selected Nuclear Materials and Engineering Systems' of Volume 11 'Ternary Alloy Systems - Phase Diagrams, Crystallographic and Thermodynamic Data critically evaluated by MSIT®' of Landolt-Börnstein - Group IV 'Physical Chemistry'. It provides data of the ternary system Oxygen-Plutonium-Uranium.

Cobalt-Free Permanent Magnet Alloys.

DTIC Science & Technology

1984-10-01

carbide co- UC CbC lumbium carbide M003 Uranium carbide - tho- UC 2 25ThC rium carbide ZrO2 MgO WOs Use of this Process for MnAlC As indicated in the...cobalt. Free World Cobal Consumption Estimated Breakdown by End Uses Magnetic alloys 20% Cemented carbides - 5% 30 SuPerolloy _ 15% Other steels and...would normally result in the formation of binary alloy of TbFe 2 and preventing the formation of amorphous alloy (Fe-B) contain- ing Tb. The

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

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

Cummins, Dustin Ray; Vogel, Sven C.; Hollis, Kendall Jon

2016-10-18

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffractionmore » data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to the process and analysis can be made, and neutron diffraction can become a viable and efficient technique for gamma/alpha phase composition determination for nuclear fuels.« less

Particle Characteristics and Densification of W6Mo5Cr4V2Co5Nb Overspray Powder

NASA Astrophysics Data System (ADS)

Pi, Ziqiang; Lu, Xin; Yang, Fei; Liu, Bowen; Jia, Chengchang; Qu, Xuanhui; Zheng, Wei; Wu, Lizhi; Shao, Qingli

2018-05-01

W6Mo5Cr4V2Co5Nb (825 K) alloy was prepared by a two-step sintering process from overspray 825 K alloy powder. The overspray powder characteristics and the microstructure and mechanical properties of the as-sintered 825 K alloy were investigated. Results showed that two types of carbides formed a network structure in the overspray powder, which had spherical or quasispherical shape: one was MC carbide that was rich in vanadium (V), and the other was M2C carbide enriched with vanadium (V) and tungsten (W). The sintered 825 K alloy contained M6C and MC carbides, of which M6C was rich in tungsten (W) and molybdenum (Mo), and both of these two carbides were uniformly distributed in the alloy matrix. The alloy had relative density of 98.43%, hardness of HRC 51.8, and superior bending strength of 2042 MPa. These mechanical properties can meet the requirements of most engineering applications.

FUSED SALT PROCESS FOR RECOVERY OF VALUES FROM USED NUCLEAR REACTOR FUELS

DOEpatents

Moore, R.H.

1960-08-01

A process is given for recovering plutonium from a neutron-irradiated uranium mass (oxide or alloy) by dissolving the mass in an about equimolar alkali metalaluminum double chloride, adding aluminum metal to the mixture obtained at a temperature of between 260 and 860 deg C, and separating a uranium-containing metal phase and a plutonium-chloride- and fission-product chloridecontaining salt phase. Dissolution can be expedited by passing carbon tetrachloride vapors through the double salt. Separation without reduction of plutonium from neutron- bombarded uranium and that of cerium from uranium are also discussed.

Use of steel and tantalum apparatus for molten Cd-Mg-Zn alloys

NASA Technical Reports Server (NTRS)

Bennett, G. A.; Burris, L., Jr.; Kyle, M. L.; Nelson, P. A.

1966-01-01

Steel and tantalum apparatus contains various ternary alloys of cadmium, zinc, and magnesium used in pyrochemical processes for the recovery of uranium-base reactor fuels. These materials exhibit good corrosion resistance at the high temperatures necessary for fuel separation in liquid metal-molten salt solvents.

2500 KW Ship Service Turbine Generator Casing Welded Inconel Plug Failure and Repair Analysis

DTIC Science & Technology

2012-06-01

tungsten arc welding process ( GTAW ) were specified. Heat input was carefully controlled. The following changes to the procedure were recommended...properties had be assigned. The turbine casing base metal is Chromium Molybdenum Alloy Steel (MIL-C-24707/2 or ASTM A217, grade WC6). The inlay is

Method for heat treating iron-nickel-chromium alloy

DOEpatents

Not Available

1980-04-03

A method is described for heat treating an age-hardenable iron-nickel-chromium alloy to obtain a morphology of the gamma-double prime phase enveloping the gamma-prime, the alloy consisting essentially of about 25 to 45% nickel, 10 to 16% chromium, 1.5 to 3% of an element selected from the group consisting of molybdenum and niobium, about 2% titanium, about 3% aluminum, and the remainder substantially all iron. To obtain optimum results, the alloy is heated to a temperature of 1025 to 1075/sup 0/C for 2 to 5 minutes, cold-worked about 20 to 60%, aged at a temperature of about 775/sup 0/C for 8 hours followed by an air-cool, and then heated to a temperature in the range of 650 to 700/sup 0/C for 2 hours followed by an air-cool.

Method for heat treating iron-nickel-chromium alloy

DOEpatents

Merrick, Howard F.; Korenko, Michael K.

1982-01-01

A method for heat treating an age-hardenable iron-nickel-chromium alloy to obtain a bimodal distribution of gamma prime phase within a network of dislocations, the alloy consisting essentially of about 25% to 45% nickel, 10% to 16% chromium, 1.5% to 3% of an element selected from the group consisting of molybdenum and niobium, about 2% titanium, about 3% aluminum, and the remainder substantially all iron. To obtain optimum results, the alloy is heated to a temperature of 1025.degree. C. to 1075.degree. C. for 2-5 minutes, cold-worked about 20% to 60%, aged at a temperature of about 775.degree. C. for 8 hours followed by an air-cool, and then heated to a temperature in the range of 650.degree. C. to 700.degree. C. for 2 hours followed by an air-cool.

Edge profiles and limiter tests in Extrap T2

NASA Astrophysics Data System (ADS)

Bergsåker, H.; Hedin, G.; Ilyinsky, L.; Larsson, D.; Möller, A.

New edge profile measurements, including calorimetric measurements of the parallel heat flux, were made in Extrap T2. Test limiters of pure molybdenum and the TZM molybdenum alloy have been exposed in the edge plasma. The surface damage was studied, mainly by microscopy. Tungsten coated graphite probes were also exposed, and the surfaces were studied by microscopy, ion beam analysis and XPS. In this case cracking and mixing of carbon and tungsten at the interface was observed in the most heated areas, whereas carbide formation at the surface was seen in less heated areas. In these tests pure Mo generally fared better than TZM, and thin and cleaner coatings fared better than thicker and less clean.

FORMING TUBES AND RODS OF URANIUM METAL BY EXTRUSION

DOEpatents

Creutz, E.C.

1959-01-27

A method and apparatus are presented for the extrusion of uranium metal. Since uranium is very brittle if worked in the beta phase, it is desirable to extrude it in the gamma phase. However, in the gamma temperature range thc uranium will alloy with the metal of the extrusion dic, and is readily oxidized to a great degree. According to this patent, uranium extrusion in thc ganmma phase may be safely carried out by preheating a billet of uranium in an inert atmosphere to a trmperature between 780 C and 1100 C. The heated billet is then placed in an extrusion apparatus having dies which have been maintained at an elevated temperature for a sufficient length of time to produce an oxide film, and placing a copper disc between the uranium billet and the die.

MOUNT ZIRKEL WILDERNESS AND VICINITY, COLORADO.

USGS Publications Warehouse

Snyder, George L.; Patten, Lowell L.

1984-01-01

Several areas of metallic and nonmetallic mineralization have been identified from surface occurrences within the Mount Zirkel Wilderness and vicinity, Colorado. Three areas of probable copper-lead-zinc-silver-gold resource potential, two areas of probable chrome-platinum resource potential, four areas of probable uranium-thorium resource potential, two areas of probable molybdenum resource potential, and one area of probable fluorspar potential were identified. No potential for fossil fuel or geothermal resources was identified.

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

Geologic report on the San Rafael Swell Drilling Project, San Rafael Swell, Utah

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

Bluhm, C.T.; Rundle, J.G.

1981-08-01

Twenty-two holes totaling 34,874 feet (10,629.6 meters) were rotary and core drilled on the northern and western flanks of the San Rafael Swell to test fluvial-lacustrine sequences of the Morrison Formation and the lower part of the Chinle Formation. The objective of the project was to obtain subsurface data so that improved uranium resource estimates could be determined for the area. Although the Brushy Basin and the Salt Wash Members of the Morrison Formation are not considered favorable in this area for the occurrence of significant uranium deposits, uranium minerals were encountered in several of the holes. Some spotty ormore » very low-grade mineralization was also encountered in the White Star Trunk area. The lower part of the Chinle Formation is considered to be favorable for potentially significant uranium deposits along the west flank of the San Rafael Swell. One hole (SR-202) east of Ferron, Utah, intersected uranium, silver, molybdenum, and copper mineralization. More exploratory drilling in the vicinity of this hole is recommended. As a result of the study of many geochemical analyses and a careful determination of the lithology shown by drilling, a sabkha environment is suggested for the concentration of uranium, zinc, iron, lead, copper, silver, and perhaps other elements in parts of the Moody Canyon Member of the Moenkopi Formation.« less

Reconnaissance of uranium and copper deposits in parts of New Mexico, Colorado, Utah, Idaho, and Wyoming

USGS Publications Warehouse

Gott, Garland B.; Erickson, Ralph L.

1952-01-01

Because of the common association of uranium and copper in several of the commercial uranium deposits in the Colorado Plateau Province, a reconnaissance was made of several known deposits of copper disseminated through sandstone to determine whether they might be a source of uranium. In order to obtain more information regarding the relationship between copper, uranium and carbonaceous materials, some of the uraniferious asphaltrite deposits in the Shinarump conglomerate along the west flank of the San Rafael Swell were also investigated briefly. During this reconnaissance 18 deposits were examined in New Mexico, eight in Utah, two in Idaho, and one each in Wyoming and Colorado. No uranium deposits of commercial grade are associated with the copper deposits that were examined. The uraniferous asphaltites in the Shinarump conglomerate of Triassic age on the west flank of the San Rafael Swell, however, are promising from the standpoint of commercial uranium production. Spectrographic analyses of crude oil, asphalt, and bituminous shales show a rather consistent suite of trace metals including vanadium, nickel, copper, cobalt, chromium, lead zinc, and molybdenum. The similarity of the metal assemblage, including uranium of the San Rafael Swell asphaltites, to the metal assemblage in crude oil and other bituminous materials suggests that these metals were concentrated in the asphaltites from petroleum. However, the hypothesis that uranium minerals were already present before the hydrocarbons were introduced and that some sort of replacement or uranium minerals by carbon compounds was effected after the petroleum migrated into the uranium deposit should not be disregarded. The widespread association of uranium with asphaltic material suggests that it also may have been concentrated by some agency connected with the formation of petroleum. The problem of the association of uranium and other trace metals with hydrocarbons should be studied further both in the field and in the laboratory.

FUEL ELEMENT AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.

1961-04-25

A nuclear fuel element in the form of a wire is reported. A bar of uranium is enclosed in a thin layer of aluminum and the composite is sheathed in beryllium, zirconium, or stainnless steel. The sheathed article is then drawn to wire form, heated to alloy the aluminum with both uranium and sheath, and finally cold worked.

Mechanical properties of weldments in experimental Fe-12Mn-0.2Ti and Fe-12Mn-1Mo-0.2Ti alloys for cryogenic service

NASA Technical Reports Server (NTRS)

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

1981-01-01

Mechanical properties of weldments in two Fe-12Mn experimental alloys designed for cryogenic service were evaluated. Weldments were made using the GTA welding process. Tests to evaluate the weldments were conducted at -196 C and included: equivalent energy fracture toughness tests; autogenous transverse weld, notched transverse weld, and longitudinal weld tensile tests; and all-weld-metal tensile tests. The Fe-12Mn-0.2Ti and Fe-12Mn-1Mo-0.2Ti alloys proved weldable for cryogenic service, with weld metal and heat-affected zone properties comparable with those of the base metal. Optimum properties were achieved in the base alloys, weld metals, and heat-affected zones after a two-step heat treatment consisting of austenitizing at 900 C followed by tempering at 500 C. The Mo-containing alloy offered a marked improvement in cryogenic properties over those of the Mo-free alloy. Molybdenum increased the amount of retained austenite and reduced the amount of epsilon martensite observed in the microstructure of the two alloys.

Development of Coatings for Tantalum Alloy Nozzle Vanes

NASA Technical Reports Server (NTRS)

Stetson, A. R.; Wimber, R. T.

1967-01-01

A group of silicide coatings developed for the T222 tantalum-base alloy have afforded over 600 hours of protection at 1600 and 2400 F during cyclic exposure in air. These coatings were applied in two steps. A modifier alloy was applied by slurry techniques and was sintered in vacuum prior to siliciding by pack cementation in argon. Application of the modifier alloy by pack cementation was found to be much less effective. The addition of titanium and vanadium to molybdenum and tungsten yielded beneficial modifier alloys, whereas the addition of chromium showed no improvement. After siliciding, the 15Ti- 35W-15V-35Mo modifier alloy exhibited the best performance; one sample survived 1064 hours of oxidation at 2400 F. This same coating was the only coating to reproducibly provide 600 hours of protection at both 1600 and 2400 F; in the second and third of three experiments, involving oxidation of three to five specimens at each temperature in each experiment, no failures were observed in 600 hours of testing. The slurry coatings were also shown to protect the Cb752 and D43 columbium-base alloys.

Radiation resistant austenitic stainless steel alloys

DOEpatents

Maziasz, P.J.; Braski, D.N.; Rowcliffe, A.F.

1987-02-11

An austenitic stainless steel alloy, with improved resistance to radiation-induced swelling and helium embrittlement, and improved resistance to thermal creep at high temperatures, consisting essentially of, by weight percent: from 16 to 18% nickel; from 13 to 17% chromium; from 2 to 3% molybdenum; from 1.5 to 2.5% manganese; from 0.01 to 0.5% silicon; from 0.2 to 0.4% titanium; from 0.1 to 0.2% niobium; from 0.1 to 0.6% vanadium; from 0.06 to 0.12% carbon; from 0.01 to 0.03% nitrogen; from 0.03 to 0.08% phosphorus; from 0.005 to 0.01% boron; and the balance iron, and wherein the alloy may be thermomechanically treated to enhance physical and mechanical properties. 4 figs.

Radiation resistant austenitic stainless steel alloys

DOEpatents

Maziasz, Philip J.; Braski, David N.; Rowcliffe, Arthur F.

1989-01-01

An austenitic stainless steel alloy, with improved resistance to radiation-induced swelling and helium embrittlement, and improved resistance to thermal creep at high temperatures, consisting essentially of, by weight percent: from 16 to 18% nickel; from 13 to 17% chromium; from 2 to 3% molybdenum; from 1.5 to 2.5% manganese; from 0.01 to 0.5% silicon; from 0.2 to 0.4% titanium; from 0.1 to 0.2% niobium; from 0.1 to 0.6% vanadium; from 0.06 to 0.12% carbon; from 0.01% to 0.03% nitrogen; from 0.03 to 0.08% phosphorus; from 0.005 to 0.01% boron; and the balance iron, and wherein the alloy may be thermomechanically treated to enhance physical and mechanical properties.

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U-Mo/Al dispersion type fuel plates

NASA Astrophysics Data System (ADS)

Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.; King, Jeffrey C.

2014-10-01

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 μm in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates.

Atomistic simulation of cubic and tetragonal phases of U-Mo alloy: Structure and thermodynamic properties

NASA Astrophysics Data System (ADS)

Starikov, S. V.; Kolotova, L. N.; Kuksin, A. Yu.; Smirnova, D. E.; Tseplyaev, V. I.

2018-02-01

We studied structure and thermodynamic properties of cubic and tetragonal phases of pure uranium and U-Mo alloys using atomistic simulations: molecular dynamics and density functional theory. The main attention was paid to the metastable γ0 -phase that is formed in U-Mo alloys at low temperature. Structure of γ0 -phase is similar to body-centered tetragonal (bct) lattice with displacement of a central atom in the basic cell along [ 001 ] direction. Such displacements have opposite orientations for part of the neighbouring basic cells. In this case, such ordering of the displacements can be designated as antiferro-displacement. Formation of such complex structure may be interpreted through forming of short U-U bonds. At heating, the tetragonal structure transforms into cubic γs -phase, still showing ordering of central atom displacements. With rise in temperature, γs -phase transforms to γ-phase with a quasi body-centered cubic (q-bcc) lattice. The local positions of uranium atoms in γ-phase correspond to γs -phase, however, orientations of the central atom displacements become disordered. Transition from γ0 to γ can be considered as antiferro-to paraelastic transition of order-disorder type. This approach to the structure description of uranium alloy allows to explain a number of unusual features found in the experiments: anisotropy of lattice at low temperature; remarkably high self-diffusion mobility in γ-phase; decreasing of electrical resistivity at heating for some alloys. In addition, important part of this work is the development of new interatomic potential for U-Mo system made with taking into account details of studied structures.

Effects of Charge Transfer on the Adsorption of CO on Small Molybdenum-Doped Platinum Clusters.

PubMed

Ferrari, Piero; Vanbuel, Jan; Tam, Nguyen Minh; Nguyen, Minh Tho; Gewinner, Sandy; Schöllkopf, Wieland; Fielicke, André; Janssens, Ewald

2017-03-23

The interaction of carbon monoxide with platinum alloy nanoparticles is an important problem in the context of fuel cell catalysis. In this work, molybdenum-doped platinum clusters have been studied in the gas phase to obtain a better understanding of the fundamental nature of the Pt-CO interaction in the presence of a dopant atom. For this purpose, Pt n + and MoPt n-1 + (n=3-7) clusters were studied by combined mass spectrometry and density functional theory calculations, making it possible to investigate the effects of molybdenum doping on the reactivity of platinum clusters with CO. In addition, IR photodissociation spectroscopy was used to measure the stretching frequency of CO molecules adsorbed on Pt n + and MoPt n-1 + (n=3-14), allowing an investigation of dopant-induced charge redistribution within the clusters. This electronic charge transfer is correlated with the observed changes in reactivity. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Fabrication of thorium bearing carbide fuels

DOEpatents

Gutierrez, Rueben L.; Herbst, Richard J.; Johnson, Karl W. R.

1981-01-01

Thorium-uranium carbide and thorium-plutonium carbide fuel pellets have been fabricated by the carbothermic reduction process. Temperatures of 1750.degree. C. and 2000.degree. C. were used during the reduction cycle. Sintering temperatures of 1800.degree. C. and 2000.degree. C. were used to prepare fuel pellet densities of 87% and >94% of theoretical, respectively. The process allows the fabrication of kilogram quantities of fuel with good reproducibility of chemicals and phase composition. Methods employing liquid techniques that form carbide microspheres or alloying-techniques which form alloys of thorium-uranium or thorium-plutonium suffer from limitation on the quantities processed of because of criticality concerns and lack of precise control of process conditions, respectively.

METHOD OF MAKING ALLOYS OF BERYLLIUM WITH PLUTONIUM AND THE LIKE

DOEpatents

Runnals, O.J.C.

1959-02-24

The production of alloys of beryllium with one or more of the metals uranium, plutonium, actinium, americium, curium, thorium, and cerium are described. A halide salt of the metal to be alloyed with the beryllium is heated at 1300 deg C in the presence of beryllium to reduce the halide to metal and cause the latter to alloy directly with the beryllium. Although the heavy metal halides are more stable, thermodynamically, than the beryllium halides, the reducing reaction proceeds to completion if the beryllium halide product is continuously removed by vacuum distillation.

High-energy, high-rate materials processing

NASA Astrophysics Data System (ADS)

Marcus, H. L.; Bourell, D. L.; Eliezer, Z.; Persad, C.; Weldon, W.

1987-12-01

The increasingly available range of pulsed-power, high energy kinetic storage devices, such as low-inductance pulse-forming networks, compulsators, and homopolar generators, is presently considered as a basis for industrial high energy/high rate (HEHR) processing to accomplish shock hardening, drilling, rapid surface alloying and melting, welding and cutting, transformation hardening, and cladding and surface melting in metallic materials. Time-temperature-transformation concepts furnish the basis for a fundamental understanding of the potential advantages of this direct pulsed power processing. Attention is given to the HEHR processing of a refractory molybdenum alloy, a nickel-base metallic glass, tungsten, titanium aluminides, and metal-matrix composites.

Some problems of brazing technology for the divertor plate manufacturing

NASA Astrophysics Data System (ADS)

Prokofiev, Yu. G.; Barabash, V. R.; Khorunov, V. F.; Maksimova, S. V.; Gervash, A. A.; Fabritsiev, S. A.; Vinokurov, V. F.

1992-09-01

Among the different design options of the ITER reactor divertor, the joints of the carbon-based materials and molybdenum alloys and joints of tungsten and copper alloys are considered. High-temperature brazing is one of the most promising joining methods for the plasma facing and heat sink materials. The use of brazing for creation of W-Cu and graphite-Mo joints are given here. In addition, the investigation results of microstructure, microhardness and mechanical properties of the joints are presented. For W-Cu samples an influence of the neutron irradiation on the joining strength was studied.

Assessment of wrought ASTM F1058 cobalt alloy properties for permanent surgical implants.

PubMed

Clerc, C O; Jedwab, M R; Mayer, D W; Thompson, P J; Stinson, J S

1997-01-01

The behavior of the ASTM F1058 wrought cobalt-chromium-nickel-molybdenum-iron alloy (commonly referred to as Elgiloy or Phynox) is evaluated in terms of mechanical properties, magnetic resonance imaging, corrosion resistance, and biocompatibility. The data found in the literature, the experimental corrosion and biocompatibility results presented in this article, and its long track record as an implant material demonstrate that the cobalt superalloy is an appropriate material for permanent surgical implants that require high yield strength and fatigue resistance combined with high elastic modulus, and that it can be safely imaged with magnetic resonance.

An evaluation of health risk to the public as a consequence of in situ uranium mining in Wyoming, USA

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

Ruedig, Elizabeth; Johnson, Thomas E.

In the United States there is considerable public concern regarding the health effects of in situ recovery uranium mining. These concerns focus principally on exposure to contaminants mobilized in groundwater by the mining process. However, the risk arising as a result of mining must be viewed in light of the presence of naturally occurring uranium ore and other constituents which comprise a latent hazard. The United States Environmental Protection Agency recently proposed new guidelines for successful restoration of an in situ uranium mine by limiting concentrations of thirteen groundwater constituents: arsenic, barium, cadmium, chromium, lead, mercury, selenium, silver, nitrate (asmore » nitrogen), molybdenum, radium, total uranium, and gross α activity. We investigated the changes occurring to these constituents at an ISR uranium mine in Wyoming, USA by comparing groundwater quality at baseline measurement to that at stability (post-restoration) testing. Of the groundwater constituents considered, only uranium and radium-226 showed significant (p < 0.05) deviation from site-wide baseline conditions in matched-wells. Uranium concentrations increased by a factor of 5.6 (95% CI 3.6–8.9 times greater) while radium-226 decreased by a factor of about one half (95% CI 0.42–0.75 times less). Change in risk was calculated using the RESRAD (onsite) code for an individual exposed as a resident-farmer; total radiation dose to a resident farmer decreased from pre-to post-mining by about 5.2 mSv y –1. As a result, higher concentrations of uranium correspond to increased biomarkers of nephrotoxicity, however the clinical significance of this increase is unclear.« less

An evaluation of health risk to the public as a consequence of in situ uranium mining in Wyoming, USA

DOE PAGES

Ruedig, Elizabeth; Johnson, Thomas E.

2015-08-30

In the United States there is considerable public concern regarding the health effects of in situ recovery uranium mining. These concerns focus principally on exposure to contaminants mobilized in groundwater by the mining process. However, the risk arising as a result of mining must be viewed in light of the presence of naturally occurring uranium ore and other constituents which comprise a latent hazard. The United States Environmental Protection Agency recently proposed new guidelines for successful restoration of an in situ uranium mine by limiting concentrations of thirteen groundwater constituents: arsenic, barium, cadmium, chromium, lead, mercury, selenium, silver, nitrate (asmore » nitrogen), molybdenum, radium, total uranium, and gross α activity. We investigated the changes occurring to these constituents at an ISR uranium mine in Wyoming, USA by comparing groundwater quality at baseline measurement to that at stability (post-restoration) testing. Of the groundwater constituents considered, only uranium and radium-226 showed significant (p < 0.05) deviation from site-wide baseline conditions in matched-wells. Uranium concentrations increased by a factor of 5.6 (95% CI 3.6–8.9 times greater) while radium-226 decreased by a factor of about one half (95% CI 0.42–0.75 times less). Change in risk was calculated using the RESRAD (onsite) code for an individual exposed as a resident-farmer; total radiation dose to a resident farmer decreased from pre-to post-mining by about 5.2 mSv y –1. As a result, higher concentrations of uranium correspond to increased biomarkers of nephrotoxicity, however the clinical significance of this increase is unclear.« less

An evaluation of health risk to the public as a consequence of in situ uranium mining in Wyoming, USA.

PubMed

Ruedig, Elizabeth; Johnson, Thomas E

2015-12-01

In the United States there is considerable public concern regarding the health effects of in situ recovery uranium mining. These concerns focus principally on exposure to contaminants mobilized in groundwater by the mining process. However, the risk arising as a result of mining must be viewed in light of the presence of naturally occurring uranium ore and other constituents which comprise a latent hazard. The United States Environmental Protection Agency recently proposed new guidelines for successful restoration of an in situ uranium mine by limiting concentrations of thirteen groundwater constituents: arsenic, barium, cadmium, chromium, lead, mercury, selenium, silver, nitrate (as nitrogen), molybdenum, radium, total uranium, and gross α activity. We investigated the changes occurring to these constituents at an ISR uranium mine in Wyoming, USA by comparing groundwater quality at baseline measurement to that at stability (post-restoration) testing. Of the groundwater constituents considered, only uranium and radium-226 showed significant (p < 0.05) deviation from site-wide baseline conditions in matched-wells. Uranium concentrations increased by a factor of 5.6 (95% CI 3.6-8.9 times greater) while radium-226 decreased by a factor of about one half (95% CI 0.42-0.75 times less). Change in risk was calculated using the RESRAD (onsite) code for an individual exposed as a resident-farmer; total radiation dose to a resident farmer decreased from pre-to post-mining by about 5.2 mSv y(-1). Higher concentrations of uranium correspond to increased biomarkers of nephrotoxicity, however the clinical significance of this increase is unclear. Published by Elsevier Ltd.

Low Mn alloy steel for cryogenic service

DOEpatents

Morris, J.W. Jr.; Niikura, M.

A ferritic cryogenic steel which has a relatively low (about 4 to 6%) manganese content and which has been made suitable for use at cryogenic temperatures by a thermal cycling treatment followed by a final tempering. The steel includes 4 to 6% manganese, 0.02 to 0.06% carbon, 0.1 to 0.4% molybdenum and 0 to 3% nickel.

78 FR 24426 - Orthopaedic and Rehabilitation Devices Panel of the Medical Devices Advisory Committee; Notice of...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-25

... application for the Kineflex/C Cervical Artificial Disc sponsored by SpinalMotion. The Kineflex/C is a metal-on-metal (cobalt chrome molybdenum alloy) cervical total disc replacement device. The Kineflex/C is... degenerative disc disease (DDD) where DDD is defined as discogenic back pain with degeneration of the disc as...

Reducing the content of alloying elements in high-speed steel during heating in salt baths

NASA Astrophysics Data System (ADS)

Kandalovskii, I. P.; Kirillov, F. F.; Dobler, V. I.

1985-07-01

A decrease in molebdenum content occurs in the surface layers during the quench heating of a tool formed from high-speed tungsten-molybdenum steel in a barium chloride salt bath after the required heating time, while a decrease in the tungsten content takes place with more prolonged hold times.

Low Mn alloy steel for cryogenic service and method of preparation

DOEpatents

Morris, Jr., John W.; Niikura, Masakazu

1981-01-01

A ferritic cryogenic steel which has a relatively low (about 4-6%) manganese content and which has been made suitable for use at cryogenic temperatures by a thermal cycling treatment followed by a final tempering. The steel includes 4-6% manganese, 0.02-0.06% carbon, 0.1-0.4% molybdenum and 0-3% nickel.

Apparatus to recover tritium from tritiated molecules

DOEpatents

Swansiger, William A.

1988-01-01

An apparatus for recovering tritium from tritiated compounds is provided, including a preheater for heating tritiated water and other co-injected tritiated compounds to temperatures of about 600.degree. C. and a reactor charged with a mixture of uranium and uranium dioxide for receiving the preheated mixture. The reactor vessel is preferably stainless steel of sufficient mass so as to function as a heat sink preventing the reactor side walls from approaching high temperatures. A disposable copper liner extends between the reaction chamber and stainless steel outer vessel to prevent alloying of the uranium with the outer vessel. The uranium dioxide functions as an insulating material and heat sink preventing the reactor side walls from attaining reaction temperatures to thereby minimize tritium permeation rates. The uranium dioxide also functions as a diluent to allow for volumetric expansion of the uranium as it is converted to uranium dioxide.

Compatibility of buffered uranium carbides with tungsten.

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1971-01-01

Results of compatibility tests between tungsten and hyperstoichiometric uranium carbide alloys run at 1800 C for 1000 and 2500 hours. These tests compared tungsten-buffered uranium carbide with tungsten-buffered uranium-zirconium carbide. The zirconium carbide addition appeared to widen the homogeneity range of the uranium carbide, making additional carbon available for reaction. Reaction layers could be formed by either of two diffusion paths, one producing UWC2, while the second resulted in the formation of W2C. UWC2 acts as a diffusion barrier for carbon and slows the growth of the reaction layer with time, while carbon diffusion is relatively rapid in W2C, allowing equilibrium to be reached in less than 2500 hours at a temperature of 1800 C.

Data Validation Package December 2015 Groundwater Sampling at the Ambrosia Lake, New Mexico, Disposal Site March 2016

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

Tsosie, Bernadette; Johnson, Dick

The Long-Term Surveillance Plan for the Ambrosia Lake, New Mexico, Disposal Site does not require groundwater monitoring because groundwater in the uppermost aquifer is of limited use, and supplemental standards have been applied to the aquifer. However, at the request of the New Mexico Environment Department, the U.S. Department of Energy conducts annual monitoring at three locations: monitoring wells 0409, 0675, and 0678. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for US. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). Monitoring Well 0409 was not sampled during this event because itmore » was dry. Water levels were measured at each sampled well. One duplicate sample was collected from location 0675. Groundwater samples from the two sampled wells were analyzed for the constituents listed in Table 1. Time-concentration graphs for selected analytes are included in this report. At well 0675, the duplicate results for total dissolved solids and for most metals (magnesium, molybdenum, potassium, selenium, sodium, and uranium) were outside acceptance criteria, which may indicate non-homogeneous conditions at this location. November 2014 results for molybdenum and uranium at well 0675 also were outside acceptance criteria. The well condition will be evaluated prior to the next sampling event.« less

Kr ion irradiation study of the depleted-uranium alloys

NASA Astrophysics Data System (ADS)

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

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

DOE PAGES

Yamamoto, Yukinori; Pint, Bruce A.; Terrani, Kurt A.; ...

2015-10-19

Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitivemore » to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741°C.« less

Effect of molybdenum additions on the microstructures and corrosion behaviours of 316L stainless steel-based alloys

DOE PAGES

Rahman, T.; Ebert, W. L.; Indacochea, J. E.

2018-02-28

Alloys were made by alloying 5, 10, 15, 17.5, and 20 wt % Mo with Type 316L stainless steel. Sigma phases containing 21–29 wt % Mo formed along the austenite grain boundaries with the addition of 5 wt % Mo and increased with additions up to 15 wt % Mo, but they decreased with further additions. Laves phases containing 33–40 wt % Mo co-precipitated at additions of 10 wt % Mo which increased with further Mo increases. The corrosion resistance, assessed by potentiodynamic polarisation in a 10 mM NaCl solution adjusted to pH 4, increased relative to Type 316L formore » alloys made with 5 and 10 wt % added Mo, but decreased with further additions due to preferential corrosion of the Laves phase. The alloy made with 10 wt % added Mo had the highest corrosion resistance due primarily to the high Mo content of the austenite.« less

Effect of molybdenum additions on the microstructures and corrosion behaviours of 316L stainless steel-based alloys

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

Rahman, T.; Ebert, W. L.; Indacochea, J. E.

Alloys were made by alloying 5, 10, 15, 17.5, and 20 wt % Mo with Type 316L stainless steel. Sigma phases containing 21–29 wt % Mo formed along the austenite grain boundaries with the addition of 5 wt % Mo and increased with additions up to 15 wt % Mo, but they decreased with further additions. Laves phases containing 33–40 wt % Mo co-precipitated at additions of 10 wt % Mo which increased with further Mo increases. The corrosion resistance, assessed by potentiodynamic polarisation in a 10 mM NaCl solution adjusted to pH 4, increased relative to Type 316L formore » alloys made with 5 and 10 wt % added Mo, but decreased with further additions due to preferential corrosion of the Laves phase. The alloy made with 10 wt % added Mo had the highest corrosion resistance due primarily to the high Mo content of the austenite.« less

Process for recovering evolved hydrogen enriched with at least one heavy hydrogen isotope

DOEpatents

Tanaka, John; Reilly, Jr., James J.

1978-01-01

This invention relates to a separation means and method for enriching a hydrogen atmosphere with at least one heavy hydrogen isotope by using a solid titaniun alloy hydride. To this end, the titanium alloy hydride containing at least one metal selected from the group consisting of vanadium, chromium, manganese, molybdenum, iron, cobalt and nickel is contacted with a circulating gaseous flow of hydrogen containing at least one heavy hydrogen isotope at a temperature in the range of -20.degree. to +40.degree. C and at a pressure above the dissociation pressure of the hydrided alloy selectively to concentrate at least one of the isotopes of hydrogen in the hydrided metal alloy. The contacting is continued until equilibrium is reached, and then the gaseous flow is isolated while the temperature and pressure of the enriched hydride remain undisturbed selectively to isolate the hydride. Thereafter, the enriched hydrogen is selectively recovered in accordance with the separation factor (S.F.) of the alloy hydride employed.

Geology and ore deposits of the Section 23 Mine, Ambrosia Lake District, New Mexico

USGS Publications Warehouse

Granger, H.C.; Santos, E.S.

1982-01-01

The section 23 mine is one of about 18 large uranium mines opened in sandstones of the fluvial Westwater Canyon Member of the Jurassic Morrison Formation in the Ambrosia Lake mining district during the early 1960s. The Ambrosia Lake district is one of several mining districts within the Grants mineral belt, an elongate zone containing many uranium deposits along the southern flank of the San Juan basin. Two distinct types of ore occur in the mine. Primary ore occurs as peneconcordant layers of uranium-rich authigenic organic matter that impregnates parts of the reduced sandstone host rocks and which are typically elongate in an east-southeast direction subparallel both to the sedimentary trends and to the present-day regional strike of the strata. These are called prefault or trend ores because of their early genesis and their elongation and alinement. A second type of ore in the mine is referred to as postfault, stacked, or redistributed ore. Its genesis was similar to that of the roll-type deposits in Tertiary rocks of Wyoming and Texas. Oxidation, related to the development of a large tongue of oxidized rock extending from Gallup to Ambrosia Lake, destroyed much of the primary ore and redistributed it as massive accumulations of lower grade ores bordering the redox interface at the edge of the tongue. Host rocks in the southern half of sec. 23 (T. 14 N., R. 10 W.) are oxidized and contain only remnants of the original, tabular, organic-rich ore. Thick bodies of roll-type ore are distributed along the leading edge of the oxidized zone, and pristine primary ore is found only near the north edge of the section. Organic matter in the primary ore was derived from humic acids that precipitated in the pores of the sandstones and fixed uranium as both coffinite and urano-organic compounds. Vanadium, molybdenum, and selenium are also associated with the ore. The secondary or roll-type ores are essentially free of organic carbon and contain uranium both as coffinite and uraninite. They also contain vanadium and selenium but are virtually devoid of molybdenum. Although much has been learned about these deposits since the time this study was conducted, in 1966, a great deal more study will by required to completely elucidate their geologic history.

Alloy Design Challenge: Development of Low Density Superalloys for Turbine Blade Applications

NASA Technical Reports Server (NTRS)

MacKay, Rebecca A.; Gabb, Timothy P.; Smialek, James L.; Nathal, Michael V.

2009-01-01

New low density single crystal (LDS) alloys have been developed for turbine blade applications, which have the potential for significant improvements in the thrust to weight ratio over current production alloys. An innovative alloying strategy was identified to achieve high temperature creep resistance, alloy density reductions, microstructural stability, and cyclic oxidation resistance. The approach relies on the use of molybdenum (Mo) as a potent solid solution strengthener for the nickel (Ni)-base superalloy; Mo has a density much closer to Ni than other refractory elements, such as rhenium (Re) or tungsten (W). A host of testing and microstructural examinations was conducted on the superalloy single crystals, including creep rupture testing, microstructural stability, cyclic oxidation, and hot corrosion. The paper will provide an overview of the single crystal properties that were generated in this new superalloy design space. The paper will also demonstrate the feasibility of this innovative approach of low density single crystal superalloy design. It will be shown that the best LDS alloy possesses the best attributes of three generations of single crystal alloys: the low density of first-generation single crystal alloys, the excellent oxidation resistance of second-generation single crystal alloys, and a creep strength which exceeds that of second and third generation alloys.

Timing of ore-related magmatism in the western Alaska Range, southwestern Alaska

USGS Publications Warehouse

Taylor, Ryan D.; Graham, Garth E.; Anderson, Eric D.; Selby, David

2014-01-01

This report presents isotopic age data from mineralized granitic plutons in an area of the Alaska Range located approximately 200 kilometers to the west-northwest of Anchorage in southwestern Alaska. Uranium-lead isotopic data and trace element concentrations of zircons were determined for 12 samples encompassing eight plutonic bodies ranging in age from approximately 76 to 57.4 millions of years ago (Ma). Additionally, a rhenium-osmium age of molybdenite from the Miss Molly molybdenum occurrence is reported (approx. 59 Ma). All of the granitic plutons in this study host gold-, copper-, and (or) molybdenum-rich prospects. These new ages modify previous interpretations regarding the age of magmatic activity and mineralization within the study area. The new ages show that the majority of the gold-quartz vein-hosting plutons examined in this study formed in the Late Cretaceous. Further work is necessary to establish the ages of ore-mineral deposition in these deposits.

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.

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

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.

Modeling and characterization of as-welded microstructure of solid solution strengthened Ni-Cr-Fe alloys resistant to ductility-dip cracking part I: Numerical modeling

NASA Astrophysics Data System (ADS)

Unfried-Silgado, Jimy; Ramirez, Antonio J.

2014-03-01

This work aims the numerical modeling and characterization of as-welded microstructure of Ni-Cr-Fe alloys with additions of Nb, Mo and Hf as a key to understand their proven resistance to ductility-dip cracking. Part I deals with as-welded structure modeling, using experimental alloying ranges and Calphad methodology. Model calculates kinetic phase transformations and partitioning of elements during weld solidification using a cooling rate of 100 K.s-1, considering their consequences on solidification mode for each alloy. Calculated structures were compared with experimental observations on as-welded structures, exhibiting good agreement. Numerical calculations estimate an increase by three times of mass fraction of primary carbides precipitation, a substantial reduction of mass fraction of M23C6 precipitates and topologically closed packed phases (TCP), a homogeneously intradendritic distribution, and a slight increase of interdendritic Molybdenum distribution in these alloys. Incidences of metallurgical characteristics of modeled as-welded structures on desirable characteristics of Ni-based alloys resistant to DDC are discussed here.

High strength nickel-chromium-iron austenitic alloy

DOEpatents

Gibson, Robert C.; Korenko, Michael K.

1980-01-01

A solid solution strengthened Ni-Cr-Fe alloy capable of retaining its strength at high temperatures and consisting essentially of 42 to 48% nickel, 11 to 13% chromium, 2.6 to 3.4% niobium, 0.2 to 1.2% silicon, 0.5 to 1.5% vanadium, 2.6 to 3.4% molybdenum, 0.1 to 0.3% aluminum, 0.1 to 0.3% titanium, 0.02 to 0.05% carbon, 0.002 to 0.015% boron, up to 0.06 zirconium, and the balance iron. After solution annealing at 1038.degree. C. for one hour, the alloy, when heated to a temperature of 650.degree. C., has a 2% yield strength of 307 MPa, an ultimate tensile strength of 513 MPa and a rupture strength of as high as 400 MPa after 100 hours.

Understanding Organic Film Behavior on Alloy and Metal Oxides

PubMed Central

Raman, Aparna; Quiñones, Rosalynn; Barriger, Lisa; Eastman, Rachel; Parsi, Arash

2010-01-01

Native oxide surfaces of stainless steel 316L and Nitinol alloys and their constituent metal oxides namely, nickel, chromium, molybdenum, manganese, iron and titanium were modified with long chain organic acids to better understand organic film formation. The adhesion and stability of films of octadecylphosphonic acid, octadecylhydroxamic acid, octadecylcarboxylic acid and octadecylsulfonic acid on these substrates was examined in this study. The films formed on these surfaces were analyzed by diffuse reflectance infrared Fourier transform spectroscopy, contact angle goniometry, atomic force microscopy and matrix assisted laser desorption ionization mass spectrometry. The effect of the acidity of the organic moiety and substrate composition on the film characteristics and stability is discussed. Interestingly, on the alloy surfaces, the presence of less reactive metal sites does not inhibit film formation. PMID:20039608

Epithermal uranium deposits in a volcanogenic context: the example of Nopal 1 deposit, Sierra de Pena Blanca, Mexico

NASA Astrophysics Data System (ADS)

Calas, G.; Angiboust, S.; Fayek, M.; Camacho, A.; Allard, T.; Agrinier, P.

2009-12-01

The Peña Blanca molybdenum-uranium field (Chihuahua, Mexico) exhibits over 100 airborne anomalies hosted in tertiary ignimbritic ash-flow tuffs (44 Ma) overlying the Pozos conglomerate and a sequence of Cretaceous carbonate rocks. Uranium occurrences are associated with breccia zones at the intersection of two or more fault systems. Periodic reactivation of these structures associated with Basin and Range and Rio Grande tectonic events resulted in the mobilization of U and other elements by meteoric fluids heated by geothermal activity. Trace element geochemistry (U, Th, REE) provides evidence for local mobilization of uranium under oxidizing conditions. In addition, O- and H-isotope geochemistry of kaolinite, smectite, opal and calcite suggests that argillic alteration proceeded at shallow depth with meteoric water at 25-75 °C. Focussed along breccia zones, fluids precipitated several generations of pyrite and uraninite together with kaolinite, as in the Nopal 1 mine, indicating that mineralization and hydrothermal alteration of volcanic tuffs are contemporaneous. Low δ34S values (~ -24.5 ‰) of pyrites intimately associated with uraninite suggest that the reducing conditions at the origin of the U-mineralization arise from biological activity. Later, the uplift of Sierra Pena Blanca resulted in oxidation and remobilization of uranium, as confirmed by the spatial distribution of radiation-induced defect centers in kaolinites. These data show that tectonism and biogenic reducing conditions can play a major role in the formation and remobilization of uranium in epithermal deposits. By comparison with the other uranium deposits at Sierra Pena Blanca and nearby Sierra de Gomez, Nopal 1 deposit is one of the few deposits having retained a reduced uranium mineralization.

Characterization of Bond Strength of U-Mo Fuel Plates Using the Laser Shockwave Technique: Capabilities and Preliminary Results

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

J. A. Smith; D. L. Cottle; B. H. Rabin

2013-09-01

This report summarizes work conducted to-date on the implementation of new laser-based capabilities for characterization of bond strength in nuclear fuel plates, and presents preliminary results obtained from fresh fuel studies on as-fabricated monolithic fuel consisting of uranium-10 wt.% molybdenum alloys clad in 6061 aluminum by hot isostatic pressing. Characterization involves application of two complementary experimental methods, laser-shock testing and laser-ultrasonic imaging, collectively referred to as the Laser Shockwave Technique (LST), that allows the integrity, physical properties and interfacial bond strength in fuel plates to be evaluated. Example characterization results are provided, including measurement of layer thicknesses, elastic properties ofmore » the constituents, and the location and nature of generated debonds (including kissing bonds). LST provides spatially localized, non-contacting measurements with minimum specimen preparation, and is ideally suited for applications involving radioactive materials, including irradiated materials. The theoretical principles and experimental approaches employed in characterizing nuclear fuel plates are described, and preliminary bond strength measurement results are discussed, with emphasis on demonstrating the capabilities and limitations of these methods. These preliminary results demonstrate the ability to distinguish bond strength variations between different fuel plates. Although additional development work is necessary to validate and qualify the test methods, these results suggest LST is viable as a method to meet fuel qualification requirements to demonstrate acceptable bonding integrity.« less

Protected Nuclear Fuel Element

DOEpatents

Kittel, J. H.; Schumar, J. F.

1962-12-01

A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

High expansion, lithium corrosion resistant sealing glasses

DOEpatents

Brow, Richard K.; Watkins, Randall D.

1991-01-01

Glass compositions containing CaO, Al.sub.2 O.sub.3, B.sub.2 O.sub.3, SrO and BaO in various combinations of mole % are provided. These compositions are capable of forming stable glass-to-metal seals with pin materials of 446 Stainless Steel and Alloy-52 rather than molybdenum, for use in harsh chemical environments, specifically in lithium batteries.

High expansion, lithium corrosion resistant sealing glasses

DOEpatents

Brow, R.K.; Watkins, R.D.

1991-06-04

Glass compositions containing CaO, Al[sub 2]O[sub 3], B[sub 2]O[sub 3], SrO and BaO in various combinations of mole % are provided. These compositions are capable of forming stable glass-to-metal seals with pin materials of 446 Stainless Steel and Alloy-52 rather than molybdenum, for use in harsh chemical environments, specifically in lithium batteries.

21 CFR 888.3380 - Hip joint femoral (hemi-hip) trunnion-bearing metal/polyacetal cemented prosthesis.

Code of Federal Regulations, 2010 CFR

2010-04-01

... device includes prostheses that consist of a metallic stem made of alloys, such as cobalt-chromium-molybdenum, with an integrated cylindrical trunnion bearing at the upper end of the stem that fits into a... head of the device to rotate on its stem. The prosthesis is intended for use with bone cement (§ 888...

21 CFR 888.3380 - Hip joint femoral (hemi-hip) trunnion-bearing metal/polyacetal cemented prosthesis.

Code of Federal Regulations, 2014 CFR

2014-04-01

... device includes prostheses that consist of a metallic stem made of alloys, such as cobalt-chromium-molybdenum, with an integrated cylindrical trunnion bearing at the upper end of the stem that fits into a... head of the device to rotate on its stem. The prosthesis is intended for use with bone cement (§ 888...

21 CFR 888.3380 - Hip joint femoral (hemi-hip) trunnion-bearing metal/polyacetal cemented prosthesis.

Code of Federal Regulations, 2011 CFR

2011-04-01

... device includes prostheses that consist of a metallic stem made of alloys, such as cobalt-chromium-molybdenum, with an integrated cylindrical trunnion bearing at the upper end of the stem that fits into a... head of the device to rotate on its stem. The prosthesis is intended for use with bone cement (§ 888...

21 CFR 888.3380 - Hip joint femoral (hemi-hip) trunnion-bearing metal/polyacetal cemented prosthesis.

Code of Federal Regulations, 2012 CFR

2012-04-01

... device includes prostheses that consist of a metallic stem made of alloys, such as cobalt-chromium-molybdenum, with an integrated cylindrical trunnion bearing at the upper end of the stem that fits into a... head of the device to rotate on its stem. The prosthesis is intended for use with bone cement (§ 888...

21 CFR 888.3380 - Hip joint femoral (hemi-hip) trunnion-bearing metal/polyacetal cemented prosthesis.

Code of Federal Regulations, 2013 CFR

2013-04-01

... device includes prostheses that consist of a metallic stem made of alloys, such as cobalt-chromium-molybdenum, with an integrated cylindrical trunnion bearing at the upper end of the stem that fits into a... head of the device to rotate on its stem. The prosthesis is intended for use with bone cement (§ 888...

Sources of ores of the ferroalloy metals

USGS Publications Warehouse

Burchard, E.F.

1933-01-01

Since all steel is made with the addition of alloying elements, the record of the metallic raw materials contributory to the steel industry would be far from complete without reference to the ferroalloy metals. This paper, therefore, supplements two preceding arvicles on the sources of our iron ores. The photographs, with the exception of those relating to molybdenum and vanadium, are by the author.

A comparison of torque expression between stainless steel, titanium molybdenum alloy, and copper nickel titanium wires in metallic self-ligating brackets.

PubMed

Archambault, Amy; Major, Thomas W; Carey, Jason P; Heo, Giseon; Badawi, Hisham; Major, Paul W

2010-09-01

The force moment providing rotation of the tooth around the x-axis (buccal-lingual) is referred to as torque expression in orthodontic literature. Many factors affect torque expression, including the wire material characteristics. This investigation aims to provide an experimental study into and comparison of the torque expression between wire types. With a worm-gear-driven torquing apparatus, wire was torqued while a bracket mounted on a six-axis load cell was engaged. Three 0.019 x 0.0195 inch wire (stainless steel, titanium molybdenum alloy [TMA], copper nickel titanium [CuNiTi]), and three 0.022 inch slot bracket combinations (Damon 3MX, In-Ovation-R, SPEED) were compared. At low twist angles (<12 degrees), the differences in torque expression between wires were not statistically significant. At twist angles over 24 degrees, stainless steel wire yielded 1.5 to 2 times the torque expression of TMA and 2.5 to 3 times that of nickel titanium (NiTi). At high angles of torsion (over 40 degrees) with a stiff wire material, loss of linear torque expression sometimes occurred. Stainless steel has the largest torque expression, followed by TMA and then NiTi.

Estimation of excess energies and activity coefficients for the penternary Ni-Cr-Co-Al-Mo system and its subsystems

NASA Astrophysics Data System (ADS)

Dogan, A.; Arslan, H.; Dogan, T.

2015-06-01

Using different prediction methods, such as the General Solution Model of Kohler and Muggianu, the excess energy and activities of molybdenum for the sections of the phase diagram for the penternary Ni-Cr-Co-Al-Mo system with mole ratios xNi/ xMo = 1, xCr/ xMo = 1, xCo/ xMo = 1, and xAl/ xMo = r = 0.5 and 1, were thermodynamically investigated at a temperature of 2000 K, whereas the excess energy and activities of Bi for the section corresponding to the ternary Bi-Ga-Sb system with mole ratio xGa/ xSb = 1/9 were thermodynamically investigated at a temperature of 1073 K. In the case of r = 0.5 and 1 in the alloys Ni-Cr-Co-Al-Mo, a positive deviation in the activity coefficient was revealed, as molybdenum content increased. Moreover, in the calculations performed in Chou's GSM model, the obtained values for excess Gibbs energies are negative in the whole concentration range of bismuth at 1073 K and exhibit the minimum of about -2.2 kJ/mol at the mole ratio xGa/ xSb = 1/9 in the alloy Bi-Ga-Sb.

China and Proliferation of Weapons of Mass Destruction and Missiles: Policy Issues

DTIC Science & Technology

2010-08-16

nuclear weapons facilities, while experts from China worked at a uranium mine at Saghand and a centrifuge facility (for uranium enrichment) near...brief interruptions.”85 84 Barbara Opall -Rome and Vago Muradian, “Bush Privately Lauds...confiscated a rare metal used to produce alloy steel (called vanadium) being smuggled to North Korea. In the same month, China’s NHI Shenyang Mining

Additive Manufacturing of Metastable Beta Titanium Alloys

NASA Astrophysics Data System (ADS)

Yannetta, Christopher J.

Additive manufacturing processes of many alloys are known to develop texture during the deposition process due to the rapid reheating and the directionality of the dissipation of heat. Titanium alloys and with respect to this study beta titanium alloys are especially susceptible to these effects. This work examines Ti-20wt%V and Ti-12wt%Mo deposited under normal additive manufacturing process parameters to examine the texture of these beta-stabilized alloys. Both microstructures contained columnar prior beta grains 1-2 mm in length beginning at the substrate with no visible equiaxed grains. This microstructure remained constant in the vanadium system throughout the build. The microstructure of the alloy containing molybdenum changed from a columnar to an equiaxed structure as the build height increased. Eighteen additional samples of the Ti-Mo system were created under different processing parameters to identify what role laser power and travel speed have on the microstructure. There appears to be a correlation in alpha lath size and power density. The two binary alloys were again deposited under the same conditions with the addition of 0.5wt% boron to investigate the effects an insoluble interstitial alloying element would have on the microstructure. The size of the prior beta grains in these two alloys were reduced with the addition of boron by approximately 50 (V) and 100 (Mo) times.

Method of making alloys of beryllium with plutonium and the like

DOEpatents

Runnals, O J.C.

1959-02-24

The production or alloys of beryllium with one or more of the metals uranium, plutonium, actinium, americium, curium, thorium, and cerium is described. A halide salt or the metal to be alloyed with the beryllium is heated at l3O0 deg C in the presence of beryllium to reduce the halide to metal and cause the latter to alloy directly with the beryllium. Although the heavy metal halides are more stable, thermodynamically, than the beryllium halides, the reducing reaction proceeds to completion if the beryllium halide product is continuously removed by vacuum distillation.

Strength of initially virgin martensites at - 196 °C after aging and tempering

NASA Astrophysics Data System (ADS)

Eldis, George T.; Cohen, Morris

1983-06-01

The compressive strength at -196°C of martensites in Fe-0.26 pct C-24 pct Ni, Fe-0.4 pct C-21 pct Ni, and Fe-0.4 pct C-18 pct Ni-3 pct Mo alloys, all with subzero M temperatures, has been determined in the virgin condition and after one hour at temperatures from -80 to +400 °C. The effects of ausforming (20 pct reduction in area of the austenite by swaging at room temperature prior to the martensitic transformation) were also investigated. For the unausformed martensites, aging at temperatures up to 0 °C results in relatively small increases in strength. Above 0 °C, the age hardening increment increases rapidly, reaching a maximum at 100 °C. Above 100 °C, the strength decreases continuously with increasing tempering temperature except for the molybdenum-containing alloy, which exhibits secondary hardening on tempering at 400 °C. For the ausformed martensites, the response to aging at subzero temperatures is greater than for unausformed material. Strength again passes through a maximum on aging at 100 °C. However, on tempering just above 100 °C, the ausformed materials show a slower rate of softening than the unausformed martensites. The strengthening produced by the ausforming treatment is largest for the Fe-0.4 pct C-18 pct Ni-3 pct Mo alloy, but there is no evidence of carbide precipitation in the deformed austenite to a°Count for this effect of molybdenum.

NORTH ABSAROKA STUDY AREA, MONTANA.

USGS Publications Warehouse

Elliott, J.E.; Stotelmeyer, R.B.

1984-01-01

A mineral survey of the North Absaroka study area in Montana was conducted. The results of this survey indicate that parts of the area are extensively mineralized and that the area has potential for resources of gold, silver, copper, molybdenum, nickel, lead, zinc, platinum-group metals, uranium, iron, manganese, chromium, tungsten, and arsenic. Six areas of probable and substantiated mineral-resource potential were identified. The nature of the geologic terrain indicates that there is little likelihood for occurrence of oil, gas, coal, or geothermal resources.

Debye temperatures and magnetic structures of UFe xAl 12- x (3.6⩽ x⩽5) intermetallic alloys

NASA Astrophysics Data System (ADS)

Rećko, K.; Dobrzyński, L.; Szymański, K.; Hoser, A.

2000-03-01

Uranium ternary compounds UFe xAl 12- x crystallize in a body-centred tetragonal structure ThMn 12 (I 4/mmm No.139). The neutron powder diffraction, magnetization measurements as well as Mössbauer investigations clearly indicate the magnetic ordering within the iron sites. The rearrangement of iron magnetic moments from uncompensated antiferromagnetic system in UFe xAl 12- x with x<4, through coexistence of antiferro- and ferromagnetic iron components (4⩽ x<5) to pure ferromagnetic ordering for alloy with x=5 is observed. The neutron diffraction studies of magnetic structures of the aforementioned powder samples show a very rich world of possible uranium-iron magnetic interactions. For all these alloys the magnetic neutron scattering is generally weak in comparison to the nuclear one. Because of identical chemical and magnetic unit cells there are no pure magnetic reflections. Therefore, in order to extract magnetic part of the scattering one should be particularly careful in taking proper account of the thermal vibration effects.

Alternative Anodes for the Electrolytic Reduction of Uranium Dioxide

NASA Astrophysics Data System (ADS)

Merwin, Augustus

Reprocessing of spent nuclear fuel is an essential step in closing the nuclear fuel cycle. In order to consume current stockpiles, ceramic uranium dioxide spent nuclear fuel will be subjected to an electrolytic reduction process. The current reduction process employs a platinum anode and a stainless steel alloy 316 cathode in a molten salt bath consisting of LiCl-2wt% Li 2O and occurs at 700°C. A major shortcoming of the existing process is the degradation of the platinum anode under the severely oxidizing conditions encountered during electrolytic reduction. This work investigates alternative anode materials for the electrolytic reduction of uranium oxide. The high temperature and extreme oxidizing conditions encountered in these studies necessitated a unique set of design constraints on the system. Thus, a customized experimental apparatus was designed and constructed. The electrochemical experiments were performed in an electrochemical reactor placed inside a furnace. This entire setup was housed inside a glove box, in order to maintain an inert atmosphere. This study investigates alternative anode materials through accelerated corrosion testing. Surface morphology was studied using scanning electron microscopy. Surface chemistry was characterized using energy dispersive spectroscopy and Raman spectroscopy. Electrochemical behavior of candidate materials was evaluated using potentiodynamic polarization characteristics. After narrowing the number of candidate electrode materials, ferrous stainless steel alloy 316, nickel based Inconel 718 and elemental tungsten were chosen for further investigation. Of these materials only tungsten was found to be sufficiently stable at the anodic potential required for electrolysis of uranium dioxide in molten salt. The tungsten anode and stainless steel alloy 316 cathode electrode system was studied at the required reduction potential for UO2 with varying lithium oxide concentrations. Electrochemical impedance spectroscopy showed mixed (kinetic and diffusion) control and an overall low impedance due to extreme corrosion. It was observed that tungsten is sufficiently stable in LiCl - 2wt% Li 2O at 700°C at the required anodic potential for the reduction of uranium oxide. This study identifies tungsten to be a superior anode material to platinum for the electrolytic reduction of uranium oxide, both in terms of superior corrosion behavior and reduced cost, and thus recommends that tungsten be further investigated as an alternative anode for the electrolytic reduction of uranium dioxide.

Reconnaissance for radioactive deposits in eastern Alaska, 1952

USGS Publications Warehouse

Nelson, Arthur Edward; West, Walter S.; Matzko, John J.

1954-01-01

Reconnaissance for radioactive deposits was conducted in selected areas of eastern Alaska during 1952. Examination of copper, silver, and molybdenum occurrences and of a reported nickel prospect in the Slana-Nabesna and Chisana districts in the eastern Alaska Range revealed a maximum radioactivity of about 0.003 percent equivalent uranium. No appreciable radioactivity anomolies were indicated by aerial and foot traverses in the area. Reconnaissance for possible lode concentrations of uranium minerals in the vicinity of reported fluoride occurrences in the Hope Creek and Miller House-Circle Hot Springs areas of the Circle quadrangle and in the Fortymile district found a maximum of 0.055 percent equivalent uranium in a float fragment of ferruginous breccia in the Hope Creek area; analysis of samples obtained in the vicinity of the other fluoride occurrences showed a maximum of only 0.005 percent equivalent uranium. No uraniferous loads were discovered in the Koyukuk-Chandalar region, nor was the source of the monazite, previously reported in the placer concentrates from the Chandalar mining district, located. The source of the uranotheorianite in the placers at Gold Bench on the South Fork of the Koyukuk River was not found during a brief reconaissance, but a placer concentrate was obtained that contains 0.18 percent equivalent uranium. This concentrate is about ten times more radioactive than concentrates previously available from the area.

Preliminary investigation of the elemental variation and diagenesis of a tabular uranium deposit, La Sal Mine, San Juan County, Utah

USGS Publications Warehouse

Brooks, Robert A.; Campbell, John A.

1976-01-01

Ore in the La Sal mine, San Juan County, Utah, occurs as a typical tabular-type uranium deposit of the-Colorado Plateau. Uranium-vanadium occurs in the Salt Wash Member of the Jurassic Morrison Formation. Chemical and petrographic analyses were used to determine elemental variation and diagenetic aspects across the orebody. Vanadium is concentrated in the dark clay matrix, which constitutes visible ore. Uranium content is greater above the vanadium zone. Calcium, carbonate carbon, and lead show greater than fifty-fold increase across the ore zone, whereas copper and organic carbon show only a several-fold increase. Large molybdenum concentrations are present in and above the tabular layer, and large selenium concentrations occur below the uranium zone within the richest vanadium zone. Iron is enriched in the vanadium horizon. Chromium is depleted from above the ore and strongly enriched below. Elements that vary directly with the vanadium content include magnesium, iron, selenium, zirconium, strontium, titanium, lead, boron, yttrium, and scandium. The diagenetic sequence is as follows: (1) formation of secondary quartz overgrowths as cement; (2) infilling and lining of remaining pores with amber opaline material; (3) formation of vanadium-rich clay matrix, which has replaced overgrowths as well as quartz grains; (4) replacement of overgrowths and detrital grains by calcite; (5) infilling of pores with barite and the introduction of pyrite and marcasite.

The structural characterization of some biomaterials, type AISI 310, used in medicine

NASA Astrophysics Data System (ADS)

Minciuna, M. G.; Vizureanu, P.; Hanganu, C.; Achitei, D. C.; Popescu, D. C.; Focsaneanu, S. C.

2016-06-01

Orthopedics biomaterials are intended for implantation in the human body and substituted or help to repair of bones, cartilage or organ transplant, and tendons. At the end of the 20th century, the availability of materials for the manufacture implants used in medicine has been the same as for other industrial applications. The most used metals for manufacturing the orthopedics implants are: stainless steels, cobalt-chrome-molybdenum alloys, titanium and his alloys. The structural researches which are made in this paper, offer a complete analysis of AISI310 stainless steels, using: optical spectrometry, X-ray diffraction and scanning electronic microscopy.

Resistance heater for use in a glass melter

DOEpatents

Routt, K.R.; Porter, M.A.

1984-01-01

A resistance heating element that includes: a resistance heating medium of a mixture of electrically conductive and insulative particles in powdered form mixed together in predetermined proportions to achieve a given resistivity; a hollow outer electrode surrounding the resistance heating medium; and an inner electrode coaxially disposed within said outer electrode. In its preferred embodiments, the electrically conductive powder is selected from the group consisting essentially of graphite, Inconel alloy, molybdenum, nichrome alloy and stainless steel, while the insulator powder is silicon dioxide or alumina. The resistance heating element, being resistant to damage from mechanical shock and corrosion at elevated temperatures, is used in a glass melter.

Effect of solutes in binary columbium /Nb/ alloys on creep strength

NASA Technical Reports Server (NTRS)

Klein, M. J.; Metcalfe, A. G.

1973-01-01

The effect of seven different solutes in binary columbium (Nb) alloys on creep strength was determined from 1400 to 3400 F for solute concentrations to 20 at.%, using a new method of creep-strength measurement. The technique permits rapid determination of approximate creep strength over a large temperature span. All of the elements were found to increase the creep strength of columbium except tantalum. This element did not strengthen columbium until the concentration exceeded 10 at.%. Hafnium, zirconium, and vanadium strengthed columbium most at low temperatures and concentrations, whereas tungsten, molybdenum, and rhenium contributed more to creep strength at high temperatures and concentrations.

Salt transport extraction of transuranium elements from LWR fuel

DOEpatents

Pierce, R.D.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Miller, W.E.

1992-11-03

A process is described for separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl[sub 2] and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750 C to about 850 C to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl[sub 2] having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO[sub 2]. The Ca metal and CaCl[sub 2] is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including MgCl[sub 2] to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy. 2 figs.

Salt transport extraction of transuranium elements from lwr fuel

DOEpatents

Pierce, R. Dean; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750.degree. C. to about 850.degree. C. to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including Mg Cl.sub.2 to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy.

Braze material for joining ceramic to metal and ceramic to ceramic surfaces and joined ceramic to metal and ceramic to ceramic article

DOEpatents

Hunt, T.K.; Novak, R.F.

1991-05-07

An improved active metal braze filler material is provided in which the coefficient of thermal expansion of the braze filler is more closely matched with that of the ceramic and metal, or two ceramics, to provide ceramic to metal, or ceramic to ceramic, sealed joints and articles which can withstand both high temperatures and repeated thermal cycling without failing. The braze filler material comprises a mixture of a material, preferably in the form of a powder, selected from the group consisting of molybdenum, tungsten, silicon carbide and mixtures thereof, and an active metal filler material selected from the group consisting of alloys or mixtures of nickel and titanium, alloys or mixtures of nickel and zirconium, alloys or mixtures of nickel, titanium, and copper, alloys or mixtures of nickel, titanium, and zirconium, alloys or mixtures of niobium and nickel, alloys or mixtures of niobium and zirconium, alloys or mixtures of niobium and titanium, alloys or mixtures of niobium, titanium, and nickel, alloys or mixtures of niobium, zirconium, and nickel, and alloys or mixtures of niobium, titanium, zirconium, and nickel. The powder component is selected such that its coefficient of thermal expansion will effect the overall coefficient of thermal expansion of the braze material so that it more closely matches the coefficients of thermal expansion of the ceramic and metal parts to be joined. 3 figures.

Braze material for joining ceramic to metal and ceramic to ceramic surfaces and joined ceramic to metal and ceramic to ceramic article

DOEpatents

Hunt, Thomas K.; Novak, Robert F.

1991-01-01

An improved active metal braze filler material is provided in which the coefficient of thermal expansion of the braze filler is more closely matched with that of the ceramic and metal, or two ceramics, to provide ceramic to metal, or ceramic to ceramic, sealed joints and articles which can withstand both high temperatures and repeated thermal cycling without failing. The braze filler material comprises a mixture of a material, preferably in the form of a powder, selected from the group consisting of molybdenum, tungsten, silicon carbide and mixtures thereof, and an active metal filler material selected from the group consisting of alloys or mixtures of nickel and titanium, alloys or mixtures of nickel and zirconium, alloys or mixtures of nickel, titanium, and copper, alloys or mixtures of nickel, titanium, and zirconium, alloys or mixtures of niobium and nickel, alloys or mixtures of niobium and zirconium, alloys or mixtures of niobium and titanium, alloys or mixtures of niobium, titanium, and nickel, alloys or mixtures of niobium, zirconium, and nickel, and alloys or mixtures of niobium, titanium, zirconium, and nickel. The powder component is selected such that its coefficient of thermal expansion will effect the overall coefficient of thermal expansion of the braze material so that it more closely matches the coefficients of thermal expansion of the ceramic and metal parts to be joined.

Design of a Fission 99 Mo Recovery Process and Implications toward Mo Adsorption Mechanism on Titania and Alumina Sorbents

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

Stepinski, Dominique C.; Youker, Amanda J.; Krahn, Elizabeth O.

2017-03-01

Molybdenum-99 is a parent of the most widely used medical isotope technetium-99m. Proliferation concerns have prompted development of alternative Mo production methods utilizing low enriched uranium. Alumina and titania sorbents were evaluated for separation of Mo from concentrated uranyl nitrate solutions. System, mass transfer, and isotherm parameters were determined to enable design of Mo separation processes under a wide range of conditions. A model-based approach was utilized to design representative commercial-scale column processes. The designs and parameters were verified with bench-scale experiments. The results are essential for design of Mo separation processes from irradiated uranium solutions, selection of support materialmore » and process optimization. Mo uptake studies show that adsorption decreases with increasing concentration of uranyl nitrate; howeveL, examination of Mo adsorption as a function of nitrate ion concentration shows no dependency, indicating that uranium competes with Mo for adsorption sites. These results are consistent with reports indicating that Mo forms inner-sphere complexes with titania and alumina surface groups.« less

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

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

Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 mu m in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the presentmore » paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates. (C) 2014 Elsevier B.V. All rights reserved.« less

Reactive Melt Infiltration Of Silicon Into Porous Carbon

NASA Technical Reports Server (NTRS)

Behrendt, Donald R.; Singh, Mrityunjay

1994-01-01

Report describes study of synthesis of silicon carbide and related ceramics by reactive melt infiltration of silicon and silicon/molybdenum alloys into porous carbon preforms. Reactive melt infiltration has potential for making components in nearly net shape, performed in less time and at lower temperature. Object of study to determine effect of initial pore volume fraction, pore size, and infiltration material on quality of resultant product.

High pressure furnace

DOEpatents

Morris, Donald E.

1993-01-01

A high temperature high pressure furnace has a hybrid partially externally heated construction. A metallic vessel fabricated from an alloy having a composition of at least 45% nickel, 15% chrome, and 10% tungsten is utilized (the preferred alloy including 55% nickel, 22% chrome, 14% tungsten, 2% molybdenum, 3% iron (maximum) and 5% cobalt (maximum). The disclosed alloy is fabricated into 11/4 or 2 inch, 32 mm or 50 mm bar stock and has a length of about 22 inches, 56 cm. This bar stock has an aperture formed therein to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the vessel is provided with a small blind aperture into which a thermocouple can be inserted. The closed end of the vessel is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior.

Computational and Experimental Studies of Microstructure-Scale Porosity in Metallic Fuels for Improved Gas Swelling Behavior

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

Mllett, Paul; McDeavitt, Sean; Deo, Chaitanya

This proposal will investigate the stability of bimodal pore size distributions in metallic uranium and uranium-zirconium alloys during sintering and re-sintering annealing treatments. The project will utilize both computational and experimental approaches. The computational approach includes both Molecular Dynamics simulations to determine the self-diffusion coefficients in pure U and U-Zr alloys in single crystals, grain boundaries, and free surfaces, as well as calculations of grain boundary and free surface interfacial energies. Phase-field simulations using MOOSE will be conducted to study pore and grain structure evolution in microstructures with bimodal pore size distributions. Experiments will also be performed to validate themore » simulations, and measure the time-dependent densification of bimodal porous compacts.« less

Processing of U-2.5Zr-7.5Nb and U-3Zr-9Nb alloys by sintering process

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

Dos Santos, A. M. M.; Ferraz, W. B.; Lameiras, F. S.

2012-07-01

To minimize the risk of nuclear proliferation, there is worldwide interest in reducing fuel enrichment of research and test reactors. To achieve this objective while still guaranteeing criticality and cycle length requirements, there is need of developing high density uranium metallic fuels. Alloying elements such as Zr, Nb and Mo are added to uranium to improve fuel performance in reactors. In this context, the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) is developing the U-2.5Zr-7.5Nb and U-3Zr-9Nb (weight %) alloys by the innovative process of sintering that utilizes raw materials in the form of powders. The powders were pressed atmore » 400 MPa and then sintered under a vacuum of about 1x10{sup -4} Torr at temperatures ranging from 1050 deg. to 1500 deg.C. The densities of the alloys were measured geometrically and by hydrostatic method and the phases identified by X ray diffraction (XRD). The microstructures of the pellets were observed by scanning electron microscopy (SEM) and the alloying elements were analyzed by energy dispersive X-ray spectroscopy (EDS). The results obtained showed the fuel density to slightly increase with the sintering temperature. The highest density achieved was approximately 80% of theoretical density. It was observed in the pellets a superficial oxide layer formed during the sintering process. (authors)« less

Structural, microstructural and thermal analysis of U-(6-x)Zr-xNb alloys (x = 0, 2, 4, 6)

NASA Astrophysics Data System (ADS)

Kaity, Santu; Banerjee, Joydipta; Parida, S. C.; Bhasin, Vivek

2018-06-01

Uranium-rich U-Zr-Nb alloy is considered as a good alternative fuel for fast reactors from the perspective of excellent dimensional stability and desired thermo-physical properties to achieve higher burnup. Detailed investigations related to the structural and microstructural characterization, thermal expansion, phase transformation, microhardness were carried out on U-6Zr, U-4Zr-2Nb, U-2Zr-4Nb and U-6Nb alloys (composition in wt%) where the total amount of alloying elements was restricted to 6 wt%. Structural, microstructural and thermal analysis studies revealed that these alloys undergo a series of transformations from high temperature bcc γ-phase to a variety of equilibrium and intermediate phases depending upon alloy composition, cooling rate and quenching. The structural analysis was carried out by Rietveld refinement. The data of U-Nb and U-Zr-Nb alloys have been highlighted and compared with binary U-Zr alloy.

Graphene coating on the surface of CoCrMo alloy enhances the adhesion and proliferation of bone marrow mesenchymal stem cells.

PubMed

Zhang, Qi; Li, Kewen; Yan, Jinhong; Wang, Zhuo; Wu, Qi; Bi, Long; Yang, Min; Han, Yisheng

2018-03-18

The objective was to investigate whether a graphene coating could improve the surface bioactivity of a cobalt-chromium-molybdenum-based alloy (CoCrMo). Graphene was produced by chemical vapor deposition and transferred to the surface of the CoCrMo alloy using an improved wet transfer approach. The morphology of the samples was observed, and the adhesion force and stabilization of graphene coating were analyzed by a nanoscratch test and ultrasonication test. In an in vitro study, the adhesion and proliferation of bone marrow mesenchymal stem cells (BMSCs) cultured on the samples were quantified via an Alamar Blue assay and cell counting kit-8 (CCK-8) assay. The results showed that it is feasible to apply graphene to modify the surface of a CoCrMo alloy, and the enhancement of the adhesion and proliferation of BMSCs was also shown in the present study. In conclusion, graphene exhibits considerable potential for enhancing the surface bioactivity of CoCrMo alloy. Copyright © 2018 Elsevier Inc. All rights reserved.

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

1993-09-01

The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

Refractory metal joining for first wall applications

NASA Astrophysics Data System (ADS)

Cadden, C. H.; Odegard, B. C.

2000-12-01

The potential use of high temperature coolant (e.g. 900°C He) in first wall structures would preclude the applicability of copper alloy heat sink materials and refractory metals would be potential replacements. Brazing trials were conducted in order to examine techniques to join tungsten armor to high tungsten (90-95 wt%) or molybdenum TZM heat sink materials. Palladium-, nickel- and zirconium-based filler metals were investigated using brazing temperatures ranging from 1000°C to 1275°C. Palladium-nickel and palladium-cobalt braze alloys were successful in producing generally sound metallurgical joints in tungsten alloy/tungsten couples, although there was an observed tendency for the pure tungsten armor material to exhibit grain boundary cracking after bonding. The zirconium- and nickel-based filler metals produced defect-containing joints, specifically cracking and porosity, respectively. The palladium-nickel braze alloy produced sound joints in the Mo TZM/tungsten couple. Substitution of a lanthanum oxide-containing, fine-grained tungsten material (for the pure tungsten) eliminated the observed tungsten grain boundary cracking.

High temperature fuel/emitter system for advanced thermionic fuel elements

NASA Astrophysics Data System (ADS)

Moeller, Helen H.; Bremser, Albert H.; Gontar, Alexander; Fiviesky, Evgeny

1997-01-01

Specialists in space applications are currently focusing on bimodal power systems designed to provide both electric power and thermal propulsion (Kennedy, 1994 and Houts, 1995). Our work showed that thermionics is a viable technology for nuclear bimodal power systems. We demonstrated that materials for a thermionic fuel-emitter combination capable of performing at operating temperatures of 2473 K are not only possible but available. The objective of this work, funded by the US Department of Energy, Office of Space and Defense Power Systems, was to evaluate the compatibility of fuel material consisting of an uranium carbide/tantalum carbide solid solution with an emitter material consisting of a monocrystalline tungsten-niobium alloy. The uranium loading of the fuel material was 70 mole% uranium carbide. The program was successfully accomplished by a B&W/SIA LUTCH team. Its workscope was integrated with tasks being performed at both Babcock & Wilcox, Lynchburg Research Center, Lynchburg, Virginia, and SIA LUTCH, Podolsk, Russia. Samples were fabricated by LUTCH and seven thermal tests were performed in a hydrogen atmosphere. The first preliminary test was performed at 2273 K by LUTCH, and the remaining six tests were performed At B&W. Three tests were performed at 2273 K, two at 2373 K, and the final test at 2473 K. The results showed that the fuel and emitter materials were compatible in the presence of hydrogen. No evidence of liquid formation, dissolution of the uranium carbide from the uranium carbide/tantalum carbide solid solution, or diffusion of the uranium into the monocrystalline tungsten alloy was observed. Among the highlights of the program was the successful export of the fuel samples from Russia and their import into the US by commercial transport. This paper will discuss the technical aspects of this work.

Radionuclides and trace metals in Canadian moose near uranium mines: comparison of radiation doses and food chain transfer with cattle and caribou.

PubMed

Thomas, Patricia; Irvine, James; Lyster, Jane; Beaulieu, Rhys

2005-05-01

Tissues from 45 moose and 4 cattle were collected to assess the health of country foods near uranium mines in northern Saskatchewan. Bone, liver, kidney, muscle and rumen contents were analyzed for uranium, radium-226 (226Ra), lead-210 (210Pb), and polonium-210 (210Po). Cesium-137 (137Cs), potassium-40 (40K), and 27 trace metals were also measured in some tissues. Within the most active mining area, Po in liver and muscle declined significantly with distance from tailings, possibly influenced by nearby natural uranium outcrops. Moose from this area had significantly higher 226Ra, 210Pb, 210Po, and 137Cs in some edible soft tissues vs. one control area. However, soil type and diet may influence concentrations as much as uranium mining activities, given that a) liver levels of uranium, 226Ra, and 210Po were similar to a second positive control area with mineral-rich shale hills and b) 210Po was higher in cattle kidneys than in all moose. Enhanced food chain transfer from rumen contents to liver was found for selenium in the main mining area and for copper, molybdenum and cadmium in moose vs. cattle. Although radiological doses to moose in the main mining area were 2.6 times higher than doses to control moose or cattle, low moose intakes yielded low human doses (0.0068 mSv y(-1)), a mere 0.3% of the dose from intake of caribou (2.4 mSv y(-1)), the dietary staple in the area.

Optimal Recycling of Steel Scrap and Alloying Elements: Input-Output based Linear Programming Method with Its Application to End-of-Life Vehicles in Japan.

PubMed

Ohno, Hajime; Matsubae, Kazuyo; Nakajima, Kenichi; Kondo, Yasushi; Nakamura, Shinichiro; Fukushima, Yasuhiro; Nagasaka, Tetsuya

2017-11-21

Importance of end-of-life vehicles (ELVs) as an urban mine is expected to grow, as more people in developing countries are experiencing increased standards of living, while the automobiles are increasingly made using high-quality materials to meet stricter environmental and safety requirements. While most materials in ELVs, particularly steel, have been recycled at high rates, quality issues have not been adequately addressed due to the complex use of automobile materials, leading to considerable losses of valuable alloying elements. This study highlights the maximal potential of quality-oriented recycling of ELV steel, by exploring the utilization methods of scrap, sorted by parts, to produce electric-arc-furnace-based crude alloy steel with minimal losses of alloying elements. Using linear programming on the case of Japanese economy in 2005, we found that adoption of parts-based scrap sorting could result in the recovery of around 94-98% of the alloying elements occurring in parts scrap (manganese, chromium, nickel, and molybdenum), which may replace 10% of the virgin sources in electric arc furnace-based crude alloy steel production.

Thermodynamic Considerations of Contamination by Alloying Elements of Remelted End-of-Life Nickel- and Cobalt-Based Superalloys

NASA Astrophysics Data System (ADS)

Lu, Xin; Matsubae, Kazuyo; Nakajima, Kenichi; Nakamura, Shinichiro; Nagasaka, Tetsuya

2016-06-01

Cobalt and nickel are high-value commodity metals and are mostly used in the form of highly alloyed materials. The alloying elements used may cause contamination problems during recycling. To ensure maximum resource efficiency, an understanding of the removability of these alloying elements and the controllability of some of the primary alloying elements is essential with respect to the recycling of end-of-life (EoL) nickel- and cobalt-based superalloys by remelting. In this study, the distribution behaviors of approximately 30 elements that are usually present in EoL nickel- and cobalt-based superalloys in the solvent metal (nickel, cobalt, or nickel-cobalt alloy), oxide slag, and gas phases during the remelting were quantitatively evaluated using a thermodynamic approach. The results showed that most of the alloying elements can be removed either in the slag phase or into the gas phase. However, the removal of copper, tin, arsenic, and antimony by remelting is difficult, and they remain as tramp elements during the recycling. On the other hand, the distribution tendencies of iron, molybdenum, and tungsten can be controlled by changing the remelting conditions. To increase the resource efficiency of recycling, preventing contamination by the tramp elements and identifying the alloying compositions of EoL superalloys are significantly essential, which will require the development of efficient prior alloy-sorting systems and advanced separation technologies.

4. DETAIL VIEW (SIDE A) OF HANDMADE STEEL BOX ASSOCIATED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

4. DETAIL VIEW (SIDE A) OF HANDMADE STEEL BOX ASSOCIATED WITH THE DEPLETED URANIUM ALLOY DEVELOPMENT AND COMPONENT FABRICATION PROCESS. THE BOX WAS USED TO TRANSFER HEATED BLOCKS OF METAL (SHOWN IN THE OPENED DOOR) FROM THE MOLTEN SALT BATH TO THE ROLLER LINES. (4/28/62) - Rocky Flats Plant, Uranium Rolling & Forming Operations, Southeast section of plant, southeast quadrant of intersection of Central Avenue & Eighth Street, Golden, Jefferson County, CO

Recovery of tritium from tritiated molecules

DOEpatents

Swansiger, William A.

1987-01-01

A method of recovering tritium from tritiated compounds comprises the steps of heating tritiated water and other co-injected tritiated compounds in a preheater to temperatures of about 600.degree. C. The mixture is injected into a reactor charged with a mixture of uranium and uranium dioxide. The injected mixture undergoes highly exothermic reactions with the uranium causing reaction temperatures to occur in excess of the melting point of uranium, and complete decomposition of the tritiated compounds to remove tritium therefrom. The uranium dioxide functions as an insulating material and heat sink preventing the reactor side walls from attaining reaction temperatures to thereby minimize tritium permeation rates. The uranium dioxide also functions as a diluent to allow for volumetric expansion of the uranium as it is converted to uranium dioxide. The reactor vessel is preferably stainless steel of sufficient mass so as to function as a heat sink preventing the reactor side walls from approaching high temperatures. A disposable copper liner extends between the reaction chamber and stainless steel outer vessel to prevent alloying of the uranium with the outer vessel. Apparatus used to carry out the method of the invention is also disclosed.