Behavior of molybdenum in mixed-oxide fuel

International Nuclear Information System (INIS)

Giacchetti, G.; Sari, C.

1976-01-01

Metallic molybdenum, Mo--Ru--Rh--Pd alloys, barium, zirconium, and tungsten were added to uranium and uranium--plutonium oxides by coprecipitation and mechanical mixture techniques. This material was treated in a thermal gradient similar to that existing in fuel during irradiation to study the behavior of molybdenum in an oxide matrix as a function of the O/(U + Pu) ratio and some added elements. Result of ceramographic and microprobe analysis shows that when the overall O/(U + Pu) ratio is less than 2, molybdenum and Mo--Ru--Rh--Pd alloy inclusions are present in the uranium--plutonium oxide matrix. If the O/(U + Pu) ratio is greater than 2, molybdenum oxidizes to MoO 2 , which is gaseous at a temperature approximately 1000 0 C. Molybdenum oxide vapor reacts with barium oxide and forms a compound that exists as a liquid phase in the columnar grain region. Molybdenum oxide also reacts with tungsten oxide (tungsten is often present as an impurity in the fuel) and forms a compound that contains approximately 40 wt percent of actinide metals. The apparent solubility of molybdenum in uranium and uranium--plutonium oxides, determined by electron microprobe, was found to be less than 250 ppM both for hypo- and hyperstoichiometric fuels

Fracture characteristics of uranium alloys by scanning electron microscopy

International Nuclear Information System (INIS)

Koger, J.W.; Bennett, R.K. Jr.

1976-10-01

The fracture characteristics of uranium alloys were determined by scanning electron microscopy. The fracture mode of stress-corrosion cracking (SCC) of uranium-7.5 weight percent niobium-2.5 weight percent zirconium (Mulberry) alloy, uranium--niobium alloys, and uranium--molybdenum alloys in aqueous chloride solutions is intergranular. The SCC fracture surface of the Mulberry alloy is characterized by very clean and smooth grain facets. The tensile-overload fracture surfaces of these alloys are characteristically ductile dimple. Hydrogen-embrittlement failures of the uranium alloys are brittle and the fracture mode is transgranular. Fracture surfaces of the uranium-0.75 weight percent titanium alloys are quasi cleavage

Effects of neutron irradiation on microstructure and deformation behaviour of mono- and polycrystalline molybdenum and its alloys

DEFF Research Database (Denmark)

Singh, B.N.; Evans, J.H.; Horsewell, A.

1998-01-01

The influence of neutron irradiation on microstructural evolution and mechanical properties of mono- and polycrystalline molybdenum and its alloys has been investigated. Tensile specimens and 3 mm diameter discs of monocrystals of pure molybdenum and Mo-5%Re were irradiated with fission neutrons...... microscopy are presented and discussed in terms of intracascade clustering of self-interstitial atoms and the role of one-dimensional glide of these clusters in controlling microstructural evolution and the resulting mechanical properties....

Progress in irradiation performance of experimental uranium - Molybdenum dispersion fuel

International Nuclear Information System (INIS)

Hofman, Gerard L.; Meyer, Mitchell K.

2002-01-01

High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL alpha-gamma hot cells. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 7 wt% and 10 wt% molybdenum. In addition, two miniplates containing solid U-10 wt% Mo foils and three containing 6 g cm -3 U 3 Si 2 are part of the test. The results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from previous tests performed to lower burnup will be presented. (author)

Recovery of uranium and molybdenum from a carbonate type uranium-molybdenum ore

International Nuclear Information System (INIS)

Zhou Genmao; Zeng Yijun; Tang Baobin; Meng Shu; Xu Guolong

2014-01-01

Based on the results of process mineralogical research of a carbonate type uranium-molybdenum ore, leaching behaviors of the uranium-molybdenum ore were studied by alkali agitation leaching, conventional alkali column leaching and alkali curing column leaching processes. The results showed that using the alkali curing column leaching process, the leaching rate of molybdenum increased to more than 90%, and the leaching rate of uranium was about 85%, Compared with the conventional alkali column leaching process, the leaching time of the alkali curing column leaching process decreased by 60 days. (authors)

Thermal cycling behaviour and thermal stability of uranium-molybdenum alloys of low molybdenum content

International Nuclear Information System (INIS)

Decours, J.; Fabrique, B.; Peault, O.

1963-01-01

We have studied the behaviour during thermal cycling of as-cast U-Mo alloys whose molybdenum content varies from 0.5 to 3 per cent; results are given concerning grain stability during extended heat treatments and the effect of treatments combining protracted heating with thermal cycling. The thermal cycling treatments were carried out at 550, 575, 600 and 625 deg C for 1000 cycles; the protracted heating experiments were done at 550, 575, 600 and 625 deg C for 2000 hours (4000 hrs at 625 deg C). The 0.5 per cent alloy resists much better to the thermal cycling than does the non-alloyed uranium. This resistance is, however, much lower than that of alloys containing over l per cent, even at 550 deg C it improves after a heat treatment for grain-refining. Alloys of over 1.1 per cent have a very good resistance to a cycling treatment even at 625 deg C, and this behaviour improves with increasing concentrations up to 3 per cent. An increase in the temperature up to the γ-phase has few disadvantages provided that it is followed by rapid cooling (50 to 100 deg C/min). The α grain is fine, the γ-phase is of the modular form, and the behaviour during a thermal cycling treatment is satisfactory. If this cooling is slow (15 deg /hr) the α-grain is coarse and cycling treatment behaviour is identical to that of the 0.5 per cent alloy. The protracted heat treatments showed that the α-grain exhibits satisfactory stability after 2000 hours at 575, 600 and 625 deg C, and after 4000 hours at 625 deg C. A heat cycling treatment carried out after these tests affects only very little the behaviour of these alloys during cycling. (authors) [fr

Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element; Etude de la trempe et du revenu a la temperature ordinaire d'alliages uranium-chrome, uranium-fer et uranium-molybdene, a faible teneur en element d'alliage

Energy Technology Data Exchange (ETDEWEB)

Delaplace, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1960-09-15

By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the {gamma} {yields} {beta} transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the {beta} {yields} {alpha} transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form {beta} at ordinary temperatures after quenching from the {beta} and {gamma} regions. The {beta} phase is particularly unstable and changes into needles of the {alpha} form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The {beta} phase obtained by quenching from the {beta} phase region is more stable than that obtained by quenching from the {gamma} region. Chromium is a more effective stabiliser of the {beta} phase than is iron. Unfortunately it causes serious surface cracking. The {beta} {yields} {alpha} transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct {gamma} {yields} {alpha} transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C

PDS 1-5. Divertor heat sink materials pre- and post-neutron irradiation. Tensile and fatigue tests of brazed joints of molybdenum alloys and 316L stainless steel

International Nuclear Information System (INIS)

Lind, Anders.

1994-01-01

Tensile specimens from brazed joints of molybdenum alloys (TZM or Mo-5%Re) and Type 316L austenitic stainless steel tubes have been tested at ambient temperature and 127 degrees C before and after neutron irradiation at about 40 degrees C to approximately 0.2 dpa. The unirradiated specimens showed generally ductile behaviour, but the irradiated specimens were notch sensitive and failed in a brittle manner with zero elongation; in all cases the fracture occurred in the molybdenum alloy. The brittle behaviour is consistent with previously published data and results from the increase in strength (radiation hardening) and the associated increase in the ductile-brittle transition temperature (radiation embrittlement) induced in the body-centered-cubic (BCC) molybdenum alloys by irradiation to relatively low displacement doses. The same type of irradiated specimens were also used in fatigue tests. However, the results from the fatigue tests are too limited and complementary studies are needed. During exposure to water locally up to 25% of the wall thickness of the Mo-alloys has corroded away. These observations cast serious doubts on the viability of the molybdenum alloys for divertor applications in fusion systems. 8 refs, 29 figs

Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element; Etude de la trempe et du revenu a la temperature ordinaire d'alliages uranium-chrome, uranium-fer et uranium-molybdene, a faible teneur en element d'alliage

Energy Technology Data Exchange (ETDEWEB)

Delaplace, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1960-09-15

By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the {gamma} {yields} {beta} transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the {beta} {yields} {alpha} transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form {beta} at ordinary temperatures after quenching from the {beta} and {gamma} regions. The {beta} phase is particularly unstable and changes into needles of the {alpha} form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The {beta} phase obtained by quenching from the {beta} phase region is more stable than that obtained by quenching from the {gamma} region. Chromium is a more effective stabiliser of the {beta} phase than is iron. Unfortunately it causes serious surface cracking. The {beta} {yields} {alpha} transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct {gamma} {yields} {alpha} transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C

Fabrication and characterisation of uranium, molybdenum, chromium, niobium and aluminium; Dobijanje i karakterizacija legura uranijuma sa molibdenom, hromom, niobijumom i aluminijumom

Energy Technology Data Exchange (ETDEWEB)

Sofrenovic, R; Isailovic, M; Kotur, Z [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-11-15

This paper describes fabrication of binary uranium alloys by melting and casting. The following alloys with nominal composition were obtained by melting in the vacuum furnace: uranium with niobium contents from 0.5%- 4.0% and uranium with molybdenum contents from 0.4% - 1.2%. Uranium alloys with chromium content from 0.4% - 1.2% and uranium alloy with 0.12% of aluminium were obtained by vacuum induction furnace (electric arc melting)

Behaviour of uranium under irradiation; Comportement de l'uranium sous irradiation

Energy Technology Data Exchange (ETDEWEB)

Adda, Y; Mustelier, J P; Quere, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

The main results obtained in a study of the formation of defects caused in uranium by fission at low temperature are reported. By irradiation at 20 K. it was possible to determine the number of Frenkel pairs produced by one fission. An analysis of the curves giving the variations in electrical resistivity shows the size of the displacement spikes and the mechanism of defect creation due to fission. Irradiations at 77 K gave additional information, showing behaviour differences in the case of recrystallised and of cold worked uranium. The diffusion of rare gases was studied using metal-rare gas alloys obtained by electrical discharge, and samples of irradiated uranium. Simple diffusion is only responsible for the release of the rare gases under vacuum in cases where the rare gas content is very low (very slightly irradiated U). On the other hand when the concentration is higher (samples prepared by electrical discharge) the gas is given off by the formation, growth and coalescence of bubbles; the apparent diffusion coefficient is then quite different from the true coefficient and cannot be used in calculations on swelling. The various factors governing the phenomenon of simple diffusion were examined. It was shown in particular that a small addition of molybdenum could reduce the diffusion coefficient by a factor of 100. The precipitation of gas in uranium (Kr), in silver (Kr) and in Al-Li alloy (He) have been followed by measurement of the crystal parameter and of the electrical resistivity, and by electron microscope examination of thin films. The important part played by dislocations in the generation and growth of bubbles has been demonstrated, and it has been shown also that precipitation of bubbles on the dislocation lattice could block the development of recrystallisation. The results of these studies were compared with observations made on the swelling of uranium and uranium alloys U Mo and U Nb strongly irradiated between 400 and 700 C. In the case of Cubic

Properties of low content uranium-molybdenum alloys which may be used as nuclear fuels

International Nuclear Information System (INIS)

Lehmann, J.; Decours, J.

1964-01-01

Metallurgical properties are given in this report of uranium-molybdenum alloys containing 0,5 to 3 per cent of molybdenum. Since some of these alloys are used in EDF power reactors are given: briefly the operating conditions imposed on nuclear fuels: maximum temperature, temperature gradient and external pressure. In the first part are considered the structural properties of the alloys correlation with the phase transformation kinetics; a description is given of the effects of certain physico-metallurgical factors on the morphology and the crystalline structure of the materials: - solidification conditions and the heredity of the γ structure, - cooling rate at the transformation points, - whether or not the intermediate γ → β transformation is suppressed In the second part we show how a knowledge of the phase transformation processes has made it possible to define the optimum preparation conditions for these materials in the form of fuel tubes intended for the EDF reactors: casting conditions, controlled cooling treatments, weldability. In the third part we study the thermal, stability during the long duration high temperature treatments and the cycles in the two zones of the diagram α + γ; β + γ the effects of the morphology (in particular the two types of α pseudo-grains observed) and of the cooling rate during the transformation point transitions are described. In the fourth part are discussed the mechanical properties: resistance to a tractive force, resistance to creep, resilience. These properties can also be affected by the γ structure heredity and by the cooling rate to which the alloy has been subjected. In conclusion we discuss the reasons which led to the choice of some of these alloys for the first EDF reactors in particular the advantages of their high creep resistance between 450 and 600 deg C for use in the form of tubes subjected to an external pressure. (authors) [fr

Set up of Uranium-Molybdenum powder production (HMD process)

International Nuclear Information System (INIS)

Lopez, Marisol; Pasqualini, Enrique E.; Gonzalez, Alfredo G.

2003-01-01

Powder metallurgy offers different alternatives for the production of Uranium-Molybdenum (UMo) alloy powder in sizes smaller than 150 microns. This powder is intended to be used as a dispersion fuel in an aluminum matrix for research, testing and radioisotopes production reactors (MTR). A particular process of massive hydriding the UMo alloy in gamma phase has been developed. This work describes the final adjustments of process variables to obtain UMo powder by hydriding-milling-de hydriding (HMD) and its capability for industrial scaling up. (author)

Production of uranium-molybdenum particles by spark-erosion

International Nuclear Information System (INIS)

Cabanillas, E.D.; Lopez, M.; Pasqualini, E.E.; Cirilo Lombardo, D.J.

2004-01-01

With the spark-erosion method we have produced spheroidal particles of an uranium-molybdenum alloy using pure water as dielectric. The particles were characterized by optical metallography, scanning electron microscopy, energy dispersive spectrometry and X-ray diffraction. Mostly spherical particles of UO 2 with a distinctive size distribution with peaks centered at 70 and 10 μm were obtained. The particles have central inclusions of U and Mo compounds

Production of uranium-molybdenum particles by spark-erosion

Energy Technology Data Exchange (ETDEWEB)

Cabanillas, E.D. E-mail: cabanill@cnea.gov.ar; Lopez, M.; Pasqualini, E.E.; Cirilo Lombardo, D.J

2004-01-01

With the spark-erosion method we have produced spheroidal particles of an uranium-molybdenum alloy using pure water as dielectric. The particles were characterized by optical metallography, scanning electron microscopy, energy dispersive spectrometry and X-ray diffraction. Mostly spherical particles of UO{sub 2} with a distinctive size distribution with peaks centered at 70 and 10 {mu}m were obtained. The particles have central inclusions of U and Mo compounds.

Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

Science.gov (United States)

Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

2014-07-01

The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

Deformation and Fracture Properties in Neutron Irradiated Pure Mo and Mo Alloys

International Nuclear Information System (INIS)

Byun, T.S.; Snead, L.; Li, M.; Cockeram, B.V.

2007-01-01

Full text of publication follows: The evolution in microstructural and mechanical properties was investigated for molybdenum and molybdenum alloys after high temperature neutron irradiation. Test materials include oxide dispersion-strengthened (ODS) molybdenum alloy, molybdenum- 0.5% titanium-0.1% zirconium (TZM) alloy, and low carbon arc-cast (LCAC) molybdenum. Tensile specimens were irradiated in high flux isotope reactor (HFIR) at temperatures in the range ∼300 - 1000 deg. C to neutron fluences of 2.28 - 24.7 x 10 25 n/m 2 (E>0.1 MeV) or 1.2-13.1 dpa. Tensile tests were performed at temperatures ranging from -150 deg. C to 1000 deg. C. To evaluate irradiation effects, true stress parameters (yield stress, plastic instability stress, and true fracture stress) and ductility parameters (uniform strain, fracture strain, and reduction area) were compared for both irradiated and non-irradiated materials. Fracture toughness was also evaluated from the fracture stress and fracture strain data using a fracture strain model. The fracture strain was used to determine the ductile-to-brittle transition temperature (DBTT). Results indicate that irradiation in the temperature range of 600 - 800 deg. C hardened the materials by up to 70%, while the irradiation hardening outside this temperature range was much lower (<40%). The plastic instability stress was strongly dependent on test temperature; however, it was nearly independent of irradiation dose and temperature. It was also found that the true fracture stress was dependent on test temperature. The true fracture stress was not significantly influenced by irradiation at elevated and high test temperatures; however, it was decreased significantly at sub-zero temperatures after irradiation due to material embrittlement. The DBTT for 600 deg. C irradiated ODS molybdenum alloy was found to be about room temperature or lower, and among the test materials the ODS alloy showed the highest resistance to irradiation embrittlement

Molybdenum from uranium solutions

International Nuclear Information System (INIS)

Gardner, H.E.

1981-01-01

A method of removing molybdenum from a uranium bearing solution is claimed. It comprises adding sufficient reactive lead compound to supply at least 90 percent of the stoichiometric quantity of lead ion required to fully react with the molybdenum present to form insoluble lead molybdate and continuing the reaction with agitation until the desired percentage of the molybdenum present has reacted with the lead ion

Evaluation of molybdenum and its alloys

International Nuclear Information System (INIS)

Lundberg, L.B.

1981-01-01

The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is critically examined. Pure molybdenum's high ductile-brittle transition temperature appears to be its major disadvantage. The candidate materials examined in detail for this application include low carbon arc-cast molybdenum, TZM-molybdenum alloy, and molybdenum-rhenium alloys. Published engineering properties are collected and compared, and it appears that Mo-Re alloys with 10 to 15% rhenium offer the best combination. Hardware is presently being made from electron beam melted Mo-13Re to test this conclusion

Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures

International Nuclear Information System (INIS)

Oliveira, Fabio Branco Vaz de

2008-01-01

This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature, time and

Internal hydrogen embrittlement of gamma-stabilized uranium alloys

International Nuclear Information System (INIS)

Powell, G.L.; Koger, J.W.; Bennett, R.K.; Williamson, A.L.; Hemperly, V.C.

1976-01-01

Relationships between the tensile ductility and fracture characteristics of as-quenched, gamma-stabilized uranium alloys (uranium--10 wt percent molybdenum, uranium--8.5 wt percent niobium, uranium--10 wt percent niobium, and uranium--7.5 wt percent niobium--2.5 wt percent zirconium), the hydrogen content of the tensile specimens, and the hydrogen gas pressure during the annealing at 850 0 C of the tensile test blanks prior to quenching were established. For these alloys, the tensile ductility decreases only slightly with increasing hydrogen content up to a critical hydrogen concentration above which the tensile ductility drops to nearly zero. The only alloy not displaying this sharp drop in tensile ductility was U--7.5 Nb--2.5 Zr, probably because sufficiently high hydrogen contents could not be achieved under our experimental arrangements. The critical hydrogen content for ductility loss increased with increasing hydrogen solubility in the alloy. Fracture surfaces produced by internal hydrogen embrittlement do not resemble those produced by stress corrosion cracking (SCC) in aqueous environments containing chloride ions. 8 figs

Kr ion irradiation study of the depleted-uranium alloys

Science.gov (United States)

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.

Separation of uranium from molybdenum by alkyl phosphoric acid extraction

Energy Technology Data Exchange (ETDEWEB)

Zhongshi, Li

1986-08-01

The regularities of separation of uranium from molybdenum by alkyl phosphoric acid extraction are described. Two parameters, i.e., density ratio of uranium to molybdenum in organic phase at first stage and density of uranium in raffinate at last stage are presented. The relationship between these parameters and purity of molybdenum and uranium products is given. The method of adjusting and controlling these parameters in experiments and production is worked out. The technical key problem in comprehensive utilization of sedimentary type uranium ore containing molybdenum with close concentration of these to elements has been solved.

Basic design of a rotating disk centrifugal atomizer for uranium-molybdenum alloys

International Nuclear Information System (INIS)

Alzari, Silvio

2001-01-01

One of the most used techniques to produce metallic powders is the centrifugal atomization with a rotating disk. This process is employ to fabricate ductile metallic particles of uranium-molybdenum alloys (typically U- 7 % Mo, by weight) for nuclear fuel elements for research and testing reactors. These alloys exhibit a face-centered cubic structure (γ phase) which is stable above 700 C degrees and can be retained at room temperature. The rotating disk centrifugal atomization allows a rapid solidification of spherical metallic droplets of about 40 to 100 μm, considered adequate to manufacture nuclear fuel elements. Besides the thermo-physical properties of both the alloy and the cooling gas, the main parameters of the process are the radius of the disk (R), the diameter of the atomization chamber (D), the disk rotation speed (ω), the liquid volume flow rate (Q) and the superheating of the liquid (ΔT). In this work, they were applied approximate analytical models to estimate the optimal geometrical and operative parameters to obtain spherical metallic powder of U- 7 % Mo alloy. Three physical phenomena were considerate: the liquid metal flow along the surface of the disk, the fragmentation and spheroidization of the droplets and the cooling and solidification of the droplets. The principal results are the more suitable gas is helium; R ≅ 20 mm; D ≥ 1 m; ≅ 20,000 - 50,000 rpm; Q ≅ 4 - 10 cm 3 /s; ΔT ≅ 100 - 200 C degrees. By applying the relevant non-dimensional parameters governing the main physical phenomena, the conclusion is that the more appropriate non-radioactive metal to simulate the atomization of U- 7 % Mo is gold [es

Kr ion irradiation study of the depleted-uranium alloys

Energy Technology Data Exchange (ETDEWEB)

Gan, J., E-mail: Jian.Gan@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Miller, B.D. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Kirk, M.A.; Rest, J. [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Allen, T.R. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

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){sub 3}, (U, Mo)(Al, Si){sub 3}, UMo{sub 2}Al{sub 20}, U{sub 6}Mo{sub 4}Al{sub 43} and UAl{sub 4}. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 {sup o}C to ion doses up to 2.5 x 10{sup 19} ions/m{sup 2} ({approx}10 dpa) with an Kr ion flux of 10{sup 16} ions/m{sup 2}/s ({approx}4.0 x 10{sup -3} dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

In-situ field-ion microscope study of the recovery behavior of heavy metal ion-irradiated tungsten, tungsten (rhenium) alloys and molybdenum

International Nuclear Information System (INIS)

Nielsen, C.H.

1977-06-01

Three field ion microscope (FIM) experiments were carried out to study the annealing behavior of heavy ion irradiated tungsten, tungsten (rhenium) alloys and molybdenum. The first experiment dealt with the stage I long-range migration of tungsten self interstitial atoms (SIAs) in high purity tungsten of resistivity ratio, R = 24,000 (R = rho 300 /rho 4 . 2 , where rho 300 and rho 4 . 2 are the room temperature and 0 0 C resistivities). The FIM specimens were irradiated in situ at 18 K with 30 keV W + ions to an average dose of 5 x 10 12 ions cm -2 and subsequently examined by the pulsed-field evaporation technique. The second experiment dealt with the phenomenon of impurity atom trapping of SIAs during long-range migration. It was shown that rhenium atoms in a tungsten matrix tend to capture tungsten SIAs and remain bound up to temperatures as high as 390 K. The final experiment was concerned with the low temperature annealing kinetics of irradiated molybdenum. High purity molybdenum of resistivity ratio R = 5700 was irradiated at 10 K with 30 keV Mo + ions to a dose of approximately 5 x 10 12 ions cm -2 . The results indicated that the electric field has only a minimal effect on the SIA annealing kinetics. This tends to strengthen the contention that the molybdenum SIA becomes mobile at 32 K

High strength tungsten heavy alloys with molybdenum additions

International Nuclear Information System (INIS)

Bose, A.; Sims, D.M.; German, R.M.

1987-01-01

Tungsten heavy alloys are candidates for numerous applications based on the unique combination of high density, high strength, and high ductility coupled with excellent machinability. Though there has been considerable research on heavy alloys, the primary focus has been on the ductility. These alloys are well suited for ballistic uses due to their high densities and it is expected that for superior ballistic performance, a high hardness, high strength and moderate ductility alloy would be ideal. The major goal of this investigation was to obtain heavy alloys with hardness greater than HRA 72. It is evident from the phase diagrams that molybdenum, which goes into solution in tungsten, nickel and iron, could act as a potential strengthening addition. With this in view, tungsten heavy alloys with molybdenum additions were fabricated from mixed elemental powders. A baseline composition of 90W-7Ni-3Fe was chosen to its good elongation and moderate strength. The molybdenum additions were made by replacing the tungsten. Compared to the baseline properties with no molybdenum addition, the strength and hardness showed a continuous increase with molybdenum addition. The ductility of the alloy continued to decrease with increasing molybdenum content, but even with 16% wt. % molybdenum of the elongation was still around 6%. An interesting facet of these alloying additions is the grain refinement that is brought about by adding to molybdenum to the system. The grain refinement is related to the lower solubility of tunbsten in the matrix due to partial displacement by molybdenum

Uranium and Molybdenum extraction from a Cerro Solo deposit ore

International Nuclear Information System (INIS)

Becquart, Elena T.; Arias, Maria J.; Fuente, Juan C. de la; Misischia, Yamila A.; Santa Cruz, Daniel E.; Tomellini, Guido C.

2009-01-01

Cerro Solo, located in Chubut, Argentina, is a sandstone type uranium-molybdenum deposit. Good recovery of both elements can be achieved by acid leaching of the ore but the presence of molybdenum in pregnant liquors is an inconvenient to uranium separation and purification. A two steps process is developed. A selective alkaline leaching of the ore with sodium hydroxide allows separating and recovering of molybdenum and after solid-liquid separation, the ore is acid leached to recover uranium. Several samples averaging 0,2% uranium and 0,1% molybdenum with variable U/Mo ratio have been used and in both steps, leaching and oxidant reagents concentration, temperature and residence time in a stirred tank leaching have been studied. In alkaline leaching molybdenum recoveries greater than 96% are achieved, with 1% uranium extraction. In acid leaching up to 93% of the uranium is extracted and Mo/U ratio in solvent extraction feed is between 0,013 and 0,025. (author)

Molybdenum-A Key Component of Metal Alloys

Science.gov (United States)

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.

Thermal stress relieving of dilute uranium alloys

International Nuclear Information System (INIS)

Eckelmeyer, K.H.

1980-01-01

The kinetics of thermal stress relieving of uranium - 2.3 wt. % niobium, uranium - 2.0 wt. % molybdenum, and uranium - 0.75 wt. % titanium are reported and discussed. Two temperature regimes of stress relieving are observed. In the low temperature regime (T 0 C) the process appears to be controlled by an athermal microplasticity mechanism which can be completely suppressed by prior age hardening. In the high temperature regime (300 0 C 0 C) the process appears to be controlled by a classical diffusional creep mechanism which is strongly dependent on temperature and time. Stress relieving is accelerated in cases where it occurs simultaneously with age hardening. The potential danger of residual stress induced stress corrosion cracking of uranium alloys is discussed. It is shown that the residual stress relief which accompanies age hardening of uranium - 0.75% titanium more than compensates for the reduction in K/sub ISCC/ caused by aging. As a result, age hardening actually decreases the susceptibility of this alloy to residual stress induced stress corrosion cracking

Alkaline elution of uranium and molybdenum and their recovery

International Nuclear Information System (INIS)

Song Wenlan; Wu Peisheng; Zhao Pinzhi; Tao Dening; Xie Chaoyan

1987-01-01

The uranium and molybdenum can be simultaneously eluted by using eluant (NH 4 ) 2 CO 3 + (NH 4 ) 2 SO 4 from resin loaded uranium and molybdenum. The ADU is precipitated from eluant by volatilization of ammonia. The molybdenum is extracted by TFA-TBP-kerosene from the filtrate at pH 3.0-3.2 with molybdenum extraction > 98%. Uranium is nearly not extracted. The precipitation of Mo is reached by sulphuric acid after stripping and the ammonium multimolybdate is obtained. This process can give the total recovery more than 99% for U and 90% for Mo. Because of the use of sulphate salt system, the hazard of NO 3 - can be avoided

Thermal stress relieving of dilute uranium alloys

International Nuclear Information System (INIS)

Eckelmeyer, K.H.

1981-01-01

The kinetics of thermal stress relieving of uranium - 2.3 wt % niobium, uranium - 2.0 wt % molybdenum, and uranium - 0.75 wt % titanium are reported and discussed. Two temperature regimes of stress relieving are observed. In the low temperature regime (T 0 C) the process appears to be controlled by an athermal microplasticity mechanism which can be completely suppressed by prior age hardening. In the high temperature regime (300 0 C 0 C) the process appears to be controlled by a classical diffusional creep mechanism which is strongly dependent on temperature and time. Stress relieving is accelerated in cases where it occurs simultaneously with age hardening. The potential danger of residual stress induced stress corrosion cracking of uranium alloys is discussed

Radiation damage of uranium

International Nuclear Information System (INIS)

Lazarevic, Dj.

1966-11-01

Study of radiation damage covered the following: Kinetics of electric resistance of uranium and uranium alloy with 1% of molybdenum dependent on the second phase and burnup rate; Study of gas precipitation and diffusion of bubbles by transmission electron microscopy; Numerical analysis of the influence of defects distribution and concentration on the rare gas precipitation in uranium; study of thermal sedimentation of uranium alloy with molybdenum; diffusion of rare gas in metal by gas chromatography method

FY16 Status Report for the Uranium-Molybdenum Fuel Concept

International Nuclear Information System (INIS)

Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.; Lavender, Curt A.; Montgomery, Robert O.; Omberg, Ronald P.; Smith, Mark T.; Webster, Ryan A.

2016-01-01

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 year 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.

FY16 Status Report for the Uranium-Molybdenum Fuel Concept

Energy Technology Data Exchange (ETDEWEB)

Bennett, Wendy D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Henager, Charles H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Montgomery, Robert O. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Mark T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Webster, Ryan A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

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 year 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.

Study of the transformation of uranium-niobium alloys with low niobium concentrations, tempered from the gamma and beta + gamma 1 regions and then annealed at different temperatures. Comparison with uranium-molybdenum alloys (1963); Etude des transformations des alliages uranium-niobium a faible teneur en niobium trempes depuis les domaines gamma et beta + gamma 1 puis revenus a differentes temperatures. Comparaison avec les alliages uranium-molybdene (1963)

Energy Technology Data Exchange (ETDEWEB)

Collot, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1962-09-15

The author shows that uranium-niobium alloys, like uranium-molybdenum alloys, tempered from the gamma region, give a martensitic phase with a structure deriving from that of alpha uranium by a slight contraction parallel to the axis [001], The critical cooling rate allowing the formation of this martensite is 80 deg. C/s at 750 deg. C. Retention of the beta phase of uranium-niobium alloys is particularly difficult, the critical retention rate being 700 deg. C/s at 668 deg. C for an alloy containing 2.5 at. per cent of Nb. This beta phase is completely converted to the alpha phase at room temperature in about 6 hours. The TTT curves of this beta alloy are effectively reduced to the lower branch of the lower 'C'. The beta phase conversion law is expressed as: 1-x = exp. (kt){sup n} x being the degree of progression of the conversion, t the time, n an exponent no-varying with temperature and having approximately the value 2 for the alloy considered, k an increasing function of temperature. The activation energy of conversion is of the order of 14,600 cal/mole. Niobium is much less active than molybdenum as a stabiliser of beta uranium. (author) [French] Dans ce travail l'auteur montre que les alliages uranium-niobium, comme d'ailleurs les alliages uranium-molybdene, trempes depuis le domaine gamma, donnent une phase martensitique dont la structure derive de celle de l'uranium alpha par une legere contraction parallele de l'axe [001]. La vitesse critique de refroidissement permettant la formation de cette martensite est de 80 deg. C/s a 750 deg. C. La retention de la phase beta des alliages uranium-niobium est particulierement delicate car la vitesse critique de retention est de 700 deg. C/s a 668 deg. C pour l'alliage a 2,5 at. pour cent de Nb. Cette phase beta se transforme completement en phase alpha a la temperature ordinaire en 6 heures environ. Les courbes TTT de cet alliage de structure beta se reduisent pratiquement a la branche inferieure du 'C' inferieur. La

Rapid analysis of molybdenum contents in molybdenum master alloys by X-ray fluorescence technique

International Nuclear Information System (INIS)

Tongkong, P.

1985-01-01

Determination of molybdenum contents in molybdenum master alloy had been performed using energy dispersive x-ray fluorescence (EDX) technique where analysis were made via standard additions and calibration curves. Comparison of EDX technique with other analyzing techniques, i.e., wavelength dispersive x-ray fluorescence, neutron activation analysis and inductive coupled plasma spectrometry, showed consistency in the results. This technique was found to yield reliable results when molybdenum contents in master alloys were in the range of 13 to 50 percent using HPGe detector or proportional counter. When the required error was set at 1%, the minimum analyzing time was found to be 30 and 60 seconds for Fe-Mo master alloys with molybdenum content of 13.54 and 49.09 percent respectively. For Al-Mo master alloys, the minimum times required were 120 and 300 seconds with molybdenum content of 15.22 and 47.26 percent respectively

Analysis of tensile and fracture toughness results on irradiated molybdenum alloys, TZM and Mo-5%Re. Analysis of results performed in the frame of the NET task PDS 1.4

Energy Technology Data Exchange (ETDEWEB)

Scibetta, M.; Chaouadi, R.; Puzzolante, J.L

1999-10-01

Due to their good resistance at high temperature, good thermal conductivity and swelling resistance, molybdenum alloys are considered amongst the candidates for divertor structural materials. However, little is known about their tensile and fracture toughness behaviour, in particular after irradiation. This report aims to investigate the tensile and fracture toughness properties of two molybdenum alloys, namely TZM and Mo-5%Re. Tensile and compact tension specimens were irradiated in the BR2 reactor at 40 and 450 degrees Celsius up to a fast neutron fluence of 3.5 1020 n/cm{sup 2} (0.2 dpa). Fracture toughness tests were performed on both precracked and notched specimens. Results show a drastic decrease of the ductility due to irradiation, but only a slight decrease of the fracture toughness in the lower shelf domain.

Analysis of tensile and fracture toughness results on irradiated molybdenum alloys, TZM and Mo-5%Re. Analysis of results performed in the frame of the NET task PDS 1.4

International Nuclear Information System (INIS)

Scibetta, M.; Chaouadi, R.; Puzzolante, J.L.

1999-10-01

Due to their good resistance at high temperature, good thermal conductivity and swelling resistance, molybdenum alloys are considered amongst the candidates for divertor structural materials. However, little is known about their tensile and fracture toughness behaviour, in particular after irradiation. This report aims to investigate the tensile and fracture toughness properties of two molybdenum alloys, namely TZM and Mo-5%Re. Tensile and compact tension specimens were irradiated in the BR2 reactor at 40 and 450 degrees Celsius up to a fast neutron fluence of 3.5 1020 n/cm 2 (0.2 dpa). Fracture toughness tests were performed on both precracked and notched specimens. Results show a drastic decrease of the ductility due to irradiation, but only a slight decrease of the fracture toughness in the lower shelf domain

Recovery of uranium from sulphate solutions containing molybdenum

International Nuclear Information System (INIS)

Weir, D.R.; Genik-Sas-Berezowsky, R.M.

1983-01-01

A process for recovering uranium from a sulphate solution containing dissolved uranium and molybdenum includes reacting the solution with ammonia (pH 8 to 10), the pH of the original solution must not exceed 5.5 and after the addition of ammonia the pH must not be in the vicinity of 7 for a significant time. The resultant uranium precipitate is relatively uncontaminated by molybdenum. The precipitate is then separated from the remaining solution while the pH is maintained within the stated range

Development of Silicide Coating on Molybdenum Alloy Cladding

International Nuclear Information System (INIS)

Lim, Woojin; Ryu, Ho Jin

2015-01-01

The molybdenum alloy is considered as one of the accident tolerant fuel (ATF) cladding materials due to its high temperature mechanical properties. However, molybdenum has a weak oxidation resistance at elevated temperatures. To modify the oxidation resistance of molybdenum cladding, silicide coating on the cladding is considered. Molybdenum silicide layers are oxidized to SiO 2 in an oxidation atmosphere. The SiO 2 protective layer isolates the substrate from the oxidizing atmosphere. Pack cementation deposition technique is widely adopted for silicide coating for molybdenum alloys due to its simple procedure, homogeneous coating quality and chemical compatibility. In this study, the pack cementation method was conducted to develop molybdenum silicide layers on molybdenum alloys. It was found that the Mo 3 Si layer was deposited on substrate instead of MoSi 2 because of short holding time. It means that through the extension of holding time, MoSi 2 layer can be formed on molybdenum substrate to enhance the oxidation resistance of molybdenum. The accident tolerant fuel (ATF) concept is to delay the process following an accident by reducing the oxidation rate at high temperatures and to delay swelling and rupture of fuel claddings. The current research for Atf can be categorized into three groups: First, modification of existing zirconium-based alloy cladding by improving the high temperature oxidation resistance and strength. Second, replacing Zirconium based alloys with alternative metallic materials such as refractory elements with high temperature oxidation resistance and strength. Third, designing alternative fuel structures using ceramic and composite systems

Experiment on bio-leaching of associated molybdenum and uranium ore

International Nuclear Information System (INIS)

Zheng Ying; Fan Baotuan; Liu Jian; Meng Yunsheng; Liu Chao

2007-01-01

Column leaching experiment results on associated molybdenum uranium ore by bacteria (T. f) are introduced. The ore are leached for 210 days using bacteria domesticated to tolerate molybdenum, the leaching of uranium is of 98% and leaching of molybdenum is of 41%. Sulphuric acid produced by bio-oxidation of sulfides in ore can meet the demand of ore leaching. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

Rest, J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)], E-mail: jrest@anl.gov; Hofman, G.L.; Kim, Yeon Soo [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2009-04-15

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 {approx}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.

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

Science.gov (United States)

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

Design of high density gamma-phase uranium alloys for LEU dispersion fuel applications

International Nuclear Information System (INIS)

Hofman, Gerard L.; Meyer, Mitchell K.; Ray, Allison E.

1998-01-01

Uranium alloys are candidates for the fuel phase in aluminium matrix dispersion fuels requiring high uranium loading. Certain uranium alloys have been shown to have good irradiation performance at intermediate burnup. previous studies have shown that acceptable fission gas swelling behavior and fuel-aluminium interaction is possible only if the fuel alloy can be maintained in the high temperature body-centered-cubic γ-phase during fabrication and irradiation, at temperatures at which αU is the equilibrium phase. transition metals in Groups V through VIII are known to allow metastable retention of the gamma phase below the equilibrium isotherm. These metals have varying degrees of effectiveness in stabilizing the gamma phase. Certain alloys are metastable for very long times at the relatively low fuel temperatures seen in research operation. In this paper, the existing data on the gamma stability of binary and ternary uranium alloys is analysed. The mechanism and kinetics of decomposition of the gamma phase are assessed with the help of metal alloy theory. Alloys with the highest possible uranium content, good gamma-phase stability, and good neutronic performance are identified for further metallurgical studies and irradiation tests. Results from theory will be compared with experimentally generated data. (author)

Development of an environmentally friendly protective coating for the depleted uranium-0.75 wt% titanium alloy

International Nuclear Information System (INIS)

Roeper, Donald F.; Chidambaram, Devicharan; Clayton, Clive R.; Halada, Gary P.; Derek Demaree, J.

2006-01-01

Molybdenum oxide-based conversion coatings have been formed on the surface of the depleted uranium-0.75 wt% titanium alloy using either concentrated nitric acid or fluorides for surface activation prior to coating formation. The acid-activated surface forms a coating that offers corrosion protection after a period of aging, when uranium species have migrated to the surface. X-ray photoelectron spectroscopy (XPS) revealed that the protective coating is primarily a polymolybdate bound to a uranyl ion. Rutherford backscattering spectroscopy (RBS) on the acid-activated coatings also shows uranium dioxide migrating to the surface. The fluoride-activated surface does not form a protective coating and there are no uranium species on the surface as indicated by XPS. The coating on the fluoride-activated samples has been found to contain a mixture of molybdenum oxides of which the main component is molybdenum trioxide and a minor component of an Mo(V) oxide

Density changes observed in pure molybdenum and Mo-41Re after irradiation in FFTF/MOTA

International Nuclear Information System (INIS)

Garner, F.A.; Greenwood, L.R.

1993-01-01

Pure molybdenum and Mo-41wt% Re, in both the 20% cold-worked and aged and the annealed and aged conditions, were irradiated in FFTF/MOTA to exposures as high as 111 dpa. Pure molybdenum appears to approach a saturation swelling level that is independent of the starting state. Cold-worked and aged molybdenum initially swells at a higher rate than that of solution annealed and aged molybdenum and overshoots the saturation level at lower irradiation temperatures. This requires that part of the accumulated swelling be removed to approach saturation, probably by void shrinkage. The alloy Mo-41Re exhibits a more complex behavior with the annealed and aged condition initially swelling faster, but eventually the density change of both conditions begins to turn downward and tends toward densification. The role of solid transmutation to Tc, Re, and Os is though to be very important in the irradiation behavior of these two metals. Calculations of transmutant generation are provided for FFTF, HFIR and STARFIRE spectra

Radiation damage of uranium; Radijaciono ostecenje urana

Energy Technology Data Exchange (ETDEWEB)

Lazarevic, Dj [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1966-11-15

Study of radiation damage covered the following: Kinetics of electric resistance of uranium and uranium alloy with 1% of molybdenum dependent on the second phase and burnup rate; Study of gas precipitation and diffusion of bubbles by transmission electron microscopy; Numerical analysis of the influence of defects distribution and concentration on the rare gas precipitation in uranium; study of thermal sedimentation of uranium alloy with molybdenum; diffusion of rare gas in metal by gas chromatography method.

Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

Energy Technology Data Exchange (ETDEWEB)

Almeida, Cirila Tacconi de

2005-07-01

The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm{sup 3} were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm{sup 3} showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

Effect of solute atoms on collision cascades in copper and molybdenum irradiated with self-ions

International Nuclear Information System (INIS)

English, C.A.; Eyre, B.L.; Wadley, H.; Stathopoulos, A.Y.

1975-01-01

An examination of the effect material purity has on the numbers and sizes of the vacancy loops formed in collision cascades produced by self-ion irradiation of copper and molybdenum is reported. It is shown that substitutional and interstitial impurities both markedly reduce the damage generated in molybdenum by 60 keV Mo + ions but little effect is seen in copper irradiated by 30 keV Cu + ions. These results are compared with recent observations of vacancy defects in type 316 stainless steel following irradiation with 40-200 keV Cr + . The comparison highlights the much lower vacancy concentration retained in visible clusters in the complex alloy

Gaseous oxygen and hydrogen embrittlements of the uranium-10 weight % molybdenum alloy

International Nuclear Information System (INIS)

Corcos, Jean.

1979-07-01

The stress corrosion of an Uranium-10 weight % Molybdenum alloy in high purity gaseous oxygen and hydrogen was studied. Tests were performed with fracture-mechanic specimens, fatigue precracked and carried out in tension with a constant sustained load. The experimental procedure enabled to determine the S.C. morphology during the test, and its kinetics. Tests in gaseous oxygen were performed with p02=0.15 MPa from 0 0 C to 100 0 C, and at 20 0 C for p02=0.15, 0.15.10 -2 and 0.15.10 -4 MPa. Two kinetic laws are proposed. Cracking is transgranular with a quasi-clivage type, and occurs on the (1 1 1) planes of the matrix. Tests in gaseous hydrogen were performed with pH2=0.15 MPa from - 50 0 C to + 135 0 C; for all the tests, even those under no exterior load, there is a failure by S.C. and macroscopic hydruration occurs. We propose a kinetic law, which may display that the hydruration phenomenon rules the S.C. propagation. We have performed the identification of the hydride, as well as the study of the precipitation. These phenomena don't occur with pH2=0.15.10 -2 MPa. The embrittlement is thought to be due to a formation-failure cycle of an hydride precipitate at the crack tip [fr

Molybdenum solubility in aluminium nitrate solutions

Energy Technology Data Exchange (ETDEWEB)

Heres, X.; Sans, D.; Bertrand, M.; Eysseric, C. [CEA, Centre de Marcoule, Nuclear Energy Division, DRCP, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Brackx, E.; Domenger, R.; Excoffier, E. [CEA, Centre de Marcoule, Nuclear Energy Division, DTEC, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Valery, J.F. [AREVA-NC, DOR/RDP, Paris - La Defense (France)

2016-07-01

For over 60 years, research reactors (RR or RTR for research testing reactors) have been used as neutron sources for research, radioisotope production ({sup 99}Mo/{sup 99m}Tc), nuclear medicine, materials characterization, etc... Currently, over 240 of these reactors are in operation in 56 countries. They are simpler than power reactors and operate at lower temperature (cooled to below 100 C. degrees). The fuel assemblies are typically plates or cylinders of uranium alloy and aluminium (U-Al) coated with pure aluminium. These fuels can be processed in AREVA La Hague plant after batch dissolution in concentrated nitric acid and mixing with UOX fuel streams. The aim of this study is to accurately measure the solubility of molybdenum in nitric acid solution containing high concentrations of aluminium. The higher the molybdenum solubility is, the more flexible reprocessing operations are, especially when the spent fuels contain high amounts of molybdenum. To be most representative of the dissolution process, uranium-molybdenum alloy and molybdenum metal powder were dissolved in solutions of aluminium nitrate at the nominal dissolution temperature. The experiments showed complete dissolution of metallic elements after 30 minutes long stirring, even if molybdenum metal was added in excess. After an induction period, a slow precipitation of molybdic acid occurs for about 15 hours. The data obtained show the molybdenum solubility decreases with increasing aluminium concentration. The solubility law follows an exponential relation around 40 g/L of aluminium with a high determination coefficient. Molybdenum solubility is not impacted by the presence of gadolinium, or by an increasing concentration of uranium. (authors)

Solid solutions of hydrogen in niobium, molybdenum and their alloys

International Nuclear Information System (INIS)

Ishikawa, T.T.

1981-01-01

The solubility of hydrogen in niobium, molybdenum and niobium-molybdenum alloys with varying atomic fraction of molybdenum from 0.15 to 0.75 was measured on the temperature range of 673 0 K to 1273 0 k for one atmosphere hydrogen pressure. The experimental technique involved the saturation of the solvent metal or alloy with hydrogen, followed by quenching and analysis of the solid solution. The results obtained of hydrogen solubility are consistent with the quasi-regular model for the dilute interstitial solid solutions. The partial molar enthalpy and partial molar entropy in excess of the dissolved hydrogen atoms were calculated from data of solubility versus reciprocal doping temperature. The variation of the relative partial molar enthalpy of hydrogen dissolved in niobium-molybdenum alloys, with the increase of molybdenum content of the alloy was analized. (Author) [pt

Extraction of molybdenum with TBP-dodecane mixture in nitric medium. Application to uranium refining

International Nuclear Information System (INIS)

Donnet, Louis

1993-01-01

Uranium ores may contain high quantities of molybdenum which represents an undesirable impurity in the uranium conversion process. Thus it is necessary to check carefully its extraction and its stripping during the purification step. The purpose of this study was to investigate the molybdenum behaviour during this step. We have developed a radiochemical method for the determination of the molybdenum concentration in each phase. This method used gamma radiation of technetium 99m issued from molybdenum 99 disintegration. After a systematic study of all the extraction parameters, we have proposed mechanisms accounting for molybdenum extraction and we have calculated the constants for the different equilibria. In particular, we have furnished new data concerning the extraction of polymerised molybdenum species with tri butyl phosphate. We have also determined the polymerization constants of molybdenum in the aqueous phase. The influence of uranium and phosphate ions on the molybdenum behaviour during the extraction and stripping has been investigated. We have shown that the extraction of molybdenum was not modified by uranium but improved in the presence of phosphate ions. In the general case, we have shown that uranium, phosphate ions and the ageing of the solvent have an unfavourable effect on stripping. We have explained this result by the evolution of the complex in the organic phase to an un-stripped polymeric form. In conclusion, our study has allowed to explain the behaviour of molybdenum during the purification process, and the role of impurities present in the industrial solutions. It would serve as a guide to improve the exploitation for a better molybdenum-uranium separation. (author) [fr

Recovery of uranium and molybdenum elements from gebel gattar raw material, eastern desert, Egypt. Vol. 3

Energy Technology Data Exchange (ETDEWEB)

El-Hazek, N T; Mahdy, M A; Mahmoud, H M.K. [Nuclear Materials Authority, Cairo, (Egypt)

1996-03-01

G. Gatter uranium mineralizations are located along the faults and fracture zones crossing G.Gattar granitic pluton and long the contact of the pluton with the hammamat sediments. Also, molybdenum id presented in more than one mode of occurrence. The molybdenum mineralization treated in this work is the dessimenated type. The uranium and molybdenum raw material was subjected to series of leaching experiments including acid and alkaline agitation, alkaline percolation, and acid heap leaching techniques. Recovery of uranium and molybdenum was achieved by anion-exchange method followed by their elution by acidified sodium chloride. Uranium precipitation was performed in the form of ammonium diuranate (Yellow Cake). On the other hand molybdenum was precipitated in the form of molybdenum oxide. A tentative flowsheet for the extraction of both uranium and molybdenum is proposed and discussed. 13 figs., 3 tabs.

Recovery of uranium and molybdenum elements from gebel gattar raw material, eastern desert, Egypt. Vol. 3

International Nuclear Information System (INIS)

El-Hazek, N.T.; Mahdy, M.A.; Mahmoud, H.M.K.

1996-01-01

G. Gatter uranium mineralizations are located along the faults and fracture zones crossing G.Gattar granitic pluton and long the contact of the pluton with the hammamat sediments. Also, molybdenum id presented in more than one mode of occurrence. The molybdenum mineralization treated in this work is the dessimenated type. The uranium and molybdenum raw material was subjected to series of leaching experiments including acid and alkaline agitation, alkaline percolation, and acid heap leaching techniques. Recovery of uranium and molybdenum was achieved by anion-exchange method followed by their elution by acidified sodium chloride. Uranium precipitation was performed in the form of ammonium diuranate (Yellow Cake). On the other hand molybdenum was precipitated in the form of molybdenum oxide. A tentative flowsheet for the extraction of both uranium and molybdenum is proposed and discussed. 13 figs., 3 tabs

Influence of Chromium and Molybdenum on the Corrosion of Nickel Based Alloys

International Nuclear Information System (INIS)

Hayes, J R; Gray, J; Szmodis, A W; Orme, C A

2005-01-01

The addition of chromium and molybdenum to nickel creates alloys with exceptional corrosion resistance in a diverse range of environments. This study examines the complementary roles of Cr and Mo in Ni alloy passivation. Four nickel alloys with varying amounts of chromium and molybdenum were studied in 1 molar salt solutions over a broad pH range. The passive corrosion and breakdown behavior of the alloys suggests that chromium is the primary element influencing general corrosion resistance. The breakdown potential was nearly independent of molybdenum content, while the repassivation potential is strongly dependant on the molybdenum content. This indicates that chromium plays a strong role in maintaining the passivity of the alloy, while molybdenum acts to stabilize the passive film after a localized breakdown event

Swelling in several commercial alloys irradiated to very high neutron fluence

International Nuclear Information System (INIS)

Gelles, D.S.; Pintler, J.S.

1984-01-01

Swelling values have been obtained from a set of commercial alloys irradiated in EBR-II to a peak fluence of 2.5 x 10 23 n/cm 2 (E > 0.1 MeV) or approx. 125 dpa covering the range 400 to 650 0 C. The alloys can be ranked for swelling resistance from highest to lowest as follows: the martensitic and ferritic alloys, the niobium based alloys, the precipitation strengthened iron and nickel based alloys, the molybdenum alloys and the austenitic alloys

Formation of uranium based nanoparticles via gamma-irradiation

Energy Technology Data Exchange (ETDEWEB)

Nenoff, Tina M., E-mail: tmnenof@sandia.gov [Nanoscale Sciences Department, Sandia National Laboratories, P.O. Box 5800, MS-1415, Albuquerque, NM 87185 (United States); Ferriera, Summer R. [Nanoscale Sciences Department, Sandia National Laboratories, P.O. Box 5800, MS-1415, Albuquerque, NM 87185 (United States); Huang, Jianyu [Center for Integrated Nanotechnology, Sandia National Laboratories, P.O. Box 5800, MS-1315, Albuquerque, NM 87185 (United States); Hanson, Donald J. [Department of Hot Cells and Gamma Facilities, Sandia National Laboratories, P.O. Box 5800, MS-1143, Albuquerque, NM 87185 (United States)

2013-11-15

Graphical abstract: TEM image of d-U nanoparticles formed in aqueous solution by gamma irradiation. Display Omitted -- Highlights: •d-U nanoparticles were grown in solution by gamma irradiation. •The reaction solution does not exceed 25 °C (room temperature). •Only after multiday exposure to air is there evidence of oxidation of the d-U nanoparticles. •Evidence of d-U alloy nanoparticle formation confirmed by TEM/energy-dispersive X-ray (EDS) analysis. -- Abstract: The ability to fabricate nuclear fuels at low temperatures allows for the production of complex Uranium metal and alloys with minimum volatility of alloy components in the process. Gamma irradiation is a valuable method for the synthesis of a wide range of metal-based nanoparticles. We report on the synthesis via room temperature radiolysis and characterization of uranium (depleted, d-U) metal and uranium–lathanide (d-ULn, Ln = lanthanide surrogates) alloy nanoparticles from aqueous acidic salt solutions. The lanthanide surrogates chosen include La and Eu due to their similarity in ionic size and charge in solution. Detailed characterization results including UV–vis, TEM/HR-TEM, and single particle EDX (elemental analyses) are presented for the room temperature formed nanoparticle products.

Formation of uranium based nanoparticles via gamma-irradiation

International Nuclear Information System (INIS)

Nenoff, Tina M.; Ferriera, Summer R.; Huang, Jianyu; Hanson, Donald J.

2013-01-01

Graphical abstract: TEM image of d-U nanoparticles formed in aqueous solution by gamma irradiation. Display Omitted -- Highlights: •d-U nanoparticles were grown in solution by gamma irradiation. •The reaction solution does not exceed 25 °C (room temperature). •Only after multiday exposure to air is there evidence of oxidation of the d-U nanoparticles. •Evidence of d-U alloy nanoparticle formation confirmed by TEM/energy-dispersive X-ray (EDS) analysis. -- Abstract: The ability to fabricate nuclear fuels at low temperatures allows for the production of complex Uranium metal and alloys with minimum volatility of alloy components in the process. Gamma irradiation is a valuable method for the synthesis of a wide range of metal-based nanoparticles. We report on the synthesis via room temperature radiolysis and characterization of uranium (depleted, d-U) metal and uranium–lathanide (d-ULn, Ln = lanthanide surrogates) alloy nanoparticles from aqueous acidic salt solutions. The lanthanide surrogates chosen include La and Eu due to their similarity in ionic size and charge in solution. Detailed characterization results including UV–vis, TEM/HR-TEM, and single particle EDX (elemental analyses) are presented for the room temperature formed nanoparticle products

Machining of uranium and uranium alloys

International Nuclear Information System (INIS)

Morris, T.O.

1981-01-01

Uranium and uranium alloys can be readily machined by conventional methods in the standard machine shop when proper safety and operating techniques are used. Material properties that affect machining processes and recommended machining parameters are discussed. Safety procedures and precautions necessary in machining uranium and uranium alloys are also covered. 30 figures

Microstructure and mechanical properties of multi-components rare earth oxide-doped molybdenum alloys

International Nuclear Information System (INIS)

Zhang Guojun; Sun Yuanjun; Zuo Chao; Wei Jianfeng; Sun Jun

2008-01-01

Pure molybdenum and molybdenum alloys doped with two- or three-components rare earth oxide particles were prepared by powder metallurgy. Both the tensile property and fracture toughness of the pure molybdenum and multi-components rare earth oxide-doped molybdenum alloys were determined at room temperature. The multi-components rare earth oxide-doped molybdenum alloys are fine grained and contain a homogeneous distribution of fine particles in the submicron and nanometer size ranges, which is why the molybdenum alloys have higher strength and fracture toughness than pure molybdenum. Quantitative analysis is used to explain the increase in yield strength with respect to grain size and second phase strengthening. Furthermore, the relationship between the tensile properties and microstructural parameters is quantitatively established

A review of chromium, molybdenum, and tungsten alloys

International Nuclear Information System (INIS)

Klopp, W.D.

1975-01-01

The mechanical properties of chromium, molybdenum, and tungsten alloys are reviewed, with particular emphasis on high-temperature strength and low-temperature ductility. Precipitate strengthening is highly effective at 0.4-0.8 Tsub(m) in these metals, with HfC being most effective in tungsten and molybdenum, and Ta(B,C) most effective in chromium. Low-temperature ductility can be improved by alloying to promote rhenium ductilizing or solution softening. The low-temperature mechanical properties of these alloys appear related to electronic interactions rather than to the usual metallurgical considerations. (Auth.)

Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

1986-09-01

Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

Evaluation of oxide dispersion strengthened (ODS) molybdenum and molybdenum-rhenium alloys

International Nuclear Information System (INIS)

Mueller, A.J.; Bianco, R.; Buckman, R.W. Jr.

1999-01-01

Oxide dispersion strengthened (ODS) molybdenum alloys being developed for high temperature applications possess excellent high temperature strength and creep resistance. In addition they exhibit a ductile-to-brittle transition temperature (DBIT) in the worked and stress-relieved condition under longitudinal tensile load well below room temperature. However, in the recrystallized condition, the DBTT maybe near or above room temperature, depending on the volume fraction of oxide dispersion and the amount of prior work. Dilute rhenium additions (7 and 14 wt.%) to ODS molybdenum were evaluated to determine their effect on low temperature ductility. The addition of 7 wt.% rhenium to the ODS molybdenum did not significantly enhance the mechanical properties. However, the addition of 14 wt.% rhenium to the ODS molybdenum resulted in a DBTT well below room temperature in both the stress-relieved and recrystallized condition. Additionally, the tensile strength of ODS Mo-14Re is greater than the base ODS molybdenum at 1,000 to 1,250 C

Special features of nickel-molybdenum alloy electrodeposition onto screen-type cathodes

International Nuclear Information System (INIS)

Aleksandrova, G.S.; Varypaev, V.N.

1982-01-01

Electrolytic nickel-molybdenum alloy, which has a rather low hydrogen overpotential and high corrosion resistance, is of interest as cathode material in industrial electrolysis. Screen-type electrodes with a nickel-molybdenum coating can be used as nonconsumable cathodes in water-activated magnesium-alloy batteries

CONCEPTUAL PROCESS DESCRIPTION FOR THE MANUFACTURE OF LOW-ENRICHED URANIUM-MOLYBDENUM FUEL

Energy Technology Data Exchange (ETDEWEB)

Daniel M. Wachs; Curtis R. Clark; Randall J. Dunavant

2008-02-01

The National Nuclear Security Agency Global Threat Reduction Initiative (GTRI) is tasked with minimizing the use of high-enriched uranium (HEU) worldwide. A key component of that effort is the conversion of research reactors from HEU to low-enriched uranium (LEU) fuels. The GTRI Convert Fuel Development program, previously known as the Reduced Enrichment for Research and Test Reactors program was initiated in 1978 by the United States Department of Energy to develop the nuclear fuels necessary to enable these conversions. The program cooperates with the research reactors’ operators to achieve this goal of HEU to LEU conversion without reduction in reactor performance. The programmatic mandate is to complete the conversion of all civilian domestic research reactors by 2014. These reactors include the five domestic high-performance research reactors (HPRR), namely: the High Flux Isotope Reactor at the Oak Ridge National Laboratory, the Advanced Test Reactor at the Idaho National Laboratory, the National Bureau of Standards Reactor at the National Institute of Standards and Technology, the Missouri University Research Reactor at the University of Missouri–Columbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for

Regularities of the vertical distribution of uranium-molybdenum mineralization

International Nuclear Information System (INIS)

Konstantinov, V.M.; Kazantsev, V.V.; Protasov, V.N.

1980-01-01

The geological structure of one of ore fields of the uranium-molybdenum formation pertaining to the northern framing of a large volcano-tectonic depression is studied. The main uranium deposits are related to necks formed by neck facies of brown liparites. Three zones are singled out within the limits of the ore field. In the upper one there are small ore bodies with a low uranium content represented by phenolite-chlorite, pitchblende 3-coffinite 3-jordizite and calcinite-sulphide associations, in the middle one - the main ore bodies formed by pitchblende 1-chlorite, molybdenite 2 (jordizite)-pitchblende 2-hydromica, coffinite 2-pyrite associations; in the lower one-thin veinlets formed by coffinite-molybdenite 1-chlorite, brannerite-pyrite and pitchblende 1-chlorite associations. Dimensions of the ore deposits depend on the neck sizes: in small necks the middle zone and, rarely, the lower one are of the industrial interest; in the large ones - the upper middle and, probably, lower ones. The regularities found can be extended to other deposits of the uranium-molybdenum formation [ru

Formation conditions for regenerated uranium blacks in uranium-molybdenum deposits

International Nuclear Information System (INIS)

Skvortsova, K.V.; Sychev, I.V.; Modnikov, I.S.; Zhil'tsova, I.G.

1980-01-01

Formation conditions of regenerated uranium blacks in the zone of incomplete oxidation and cementation of uranium-molybdenum deposit have been studied. Mixed and regenerated blacks were differed from residual ones by the method of determining excess quantity of lead isotope (Pb 206 ) in ores. Determined were the most favourable conditions for formation of regenerated uranium blacks: sheets of brittle and permeable volcanic rocks characterized by heterogeneous structure of a section, by considerable development of gentle interlayer strippings and zones of hydrothermal alteration; predominance of reduction conditions in a media over oxidation ones under limited oxygen access and other oxidating agents; the composition of hypogenic ores characterized by optimum correlations of uranium minerals, sulfides and carbonates affecting violations of pH in oxidating solutions in the range of 5-6; the initial composition of ground water resulting from climatic conditions of the region and the composition of ore-bearing strata and others. Conditions unfavourable for the formation of regenerated uranium blacks are shown

A study on direct alloying with molybdenum oxides by feed wire method

Directory of Open Access Journals (Sweden)

Jingjing Zou

2018-04-01

Full Text Available Direct alloying with molybdenum oxides has been regarded in years; the main addition methods are adding to the bottom of electric arc furnace (EAF with scrap, adding to the ladle during the converter tapping and mixing molybdenum oxide, lime and reductant to prepare pellet added to basic oxygen furnace (BOF. In this paper, a new method for direct alloying with molybdenum trioxide is proposed, adding molybdenum trioxide molten steel by feeding wire method in ladle furnace (LF refining process. The feasibility of molybdenum oxide reduction, the influence rules of bottom-blown on liquid steel fluidity and the yield of molybdenum by feeding wire method were analyzed. Results show that molybdenum oxide can be reduced by [Al], [Si], [C], and even [Fe] in molten steel. Bottom blowing position has a significant influence on the flow of molten steel when the permeable brick is located in 1/2 radius. The yields of Mo are higher than 97% for the experiments with feed wire method, the implementation of direct alloying with molybdenum trioxide by feed wire method works even better than that uses of ferromolybdenum in the traditional process.

Mechanical properties of Mo and TZM alloy neutron-irradiated at high temperatures

International Nuclear Information System (INIS)

Ueda, Kazukiyo; Satou, Manabu; Hasegawa, Akira; Abe, Katsunori

1997-01-01

This work reports the mechanical properties of irradiated molybdenum (Mo) and its alloy, TZM. Recrystallized and stress-relieved specimens were irradiated at five temperatures between 373 and 800degC in FFTF/MOTA to fluence levels of 6.8 to 34 dpa. Irradiation embrittlement and hardening were evaluated by three-point bend test and Vickers hardness test, respectively. Stress-relieved materials showed the enough ductility even after high fluence irradiation. The role of layered structure of stress-relieved specimen was discussed. (author)

Strengthening and elongation mechanism of Lanthanum-doped Titanium-Zirconium-Molybdenum alloy

Energy Technology Data Exchange (ETDEWEB)

Hu, Ping, E-mail: huping1985@126.com [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Jinduicheng Molybdenum Co., Ltd., Xi’an 710068 (China); Hu, Bo-liang; Wang, Kuai-she; Song, Rui; Yang, Fan [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Yu, Zhi-tao [Ruifulai Tungsten & Molybdenum Co., Ltd., Xi’an 721914 (China); Tan, Jiang-fei [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Cao, Wei-cheng; Liu, Dong-xin; An, Geng [Jinduicheng Molybdenum Co., Ltd., Xi’an 710068 (China); Guo, Lei [Ruifulai Tungsten & Molybdenum Co., Ltd., Xi’an 721914 (China); Yu, Hai-liang [School of Mechanical, Materials and Mechatronics Engineering, University of Wollongong, NSW 2522 (Australia)

2016-12-15

The microstructural contributes to understand the strengthening and elongation mechanism in Lanthanum-doped Titanium-Zirconium-Molybdenum alloy. Lanthanum oxide particles not only act as heterogeneous nucleation core, but also act as the second phase to hinder the grain growth during sintering crystallization. The molybdenum substrate formed sub-grain under the effect of second phase when the alloy rolled to plate.

Critical evaluation of molybdenum and its alloys for use in space reactor core heat pipes

International Nuclear Information System (INIS)

Lundberg, L.B.

1981-01-01

The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is examined, and the advantages and disadvantages of this material are brought into focus. Even though pure molybdenum heat pipes have been built and tested, this metal's high ductile-brittle transition temperature and modest creep strength place significant design restrictions on a core heat pipe made from it. Molybdenum alloys are examined with regard to their promise as potential replacements for pure molybdenum. The properties of TZM and molybdenum-rhenium alloys are examined, and it appears that Mo-Re alloys with 10 to 15 wt % rhenium offer the most advantage as an alternative to pure molybdenum in space reactor core heat pipes

Utilization of epithermal neutrons for the determination of molybdenum in the presence of uranium

International Nuclear Information System (INIS)

Oliveira Melo, M.A.M. de.

1984-05-01

Activation analysis by means of selective activation with epithermal neutrons is proposed for the determination of molybdenum in samples when uranium is present. Instrumental activation analysis with epithermal neutrons is advantageous for the determination of elements with large resonance integral, as compared to its thermal neutron activation cross section. The main reason for using this method is the serious interference caused by 99 Mo produced by fission of 235 U. This effect is strongly reduced by using the epicadmium irradiation technique. The filter efficiency has been investigated by irradiation experiments with bare and cadmium-covered samples. A solvent extraction process for uranium, before irradiation, is proposed to reduce sample background. The determination of Mo in leach samples is proposed in order to support the analytical needs of Figueira and Pocos de Caldas Mineral Prospection Programme of Departamento de Tecnologia Mineral from CDTN/NUCLEBRAS (MG,Brazil). The introduction of activation analysis with epithermal neutrons as a routine analytical tool in CDTN is our main goal. This method represents one more opportunity for exploring the analytical facilities available at TRIGA MARK I IPR-R1 nuclear reactor. (Author) [pt

On the mechanism of dispersion hardening in molybdenum-carbide alloy systems

International Nuclear Information System (INIS)

Shulepov, V.I.; Yudkovskij, S.I.; Batenina, O.I. et al.

1975-01-01

The effect of heat treatment of the forming alloys of the Mo-Ti-C and Mo-Ti-Zr-C systems (at the temperatures below the recrystallization temperature) on the structure, distribution of carbon and mechanical properties of the alloys is studied. It is shown that the dispersion-strengthened state of the molybdenum alloys may be obtained on the account of the deformation ageing effect, rather than through the use of the standard heat-treatment procedure (hardening plus ageing). On the basis of the experimental results a theoretical explanation of strengthening of the high-alloy molybdenum-titanum-carbon system is given

β → α isothermal transformation in pure and weakly alloyed uranium

International Nuclear Information System (INIS)

Aubert, H.; Lelong, C.

1966-01-01

The TTT diagrams describing the β → α isothermal transformation have been made by isothermal dilatometry for pure uranium and 21 alloys based on chromium, silicon, molybdenum, iron, aluminium, zirconium. The thermal cycle preceding the isothermal step influences the decomposition kinetics at temperature corresponding to the eutectoid and martensitic mechanisms, but not in the range where the bainitic transformation occurs. The stability of the β phase decreases with the chromium, molybdenum and silicon concentration: it is affected differently for each of the three transformation mechanisms. The ternary additions, even at very low concentration have a considerable effect on the stability. When the concentration decreases the martensitic mechanism is active at progressively higher temperature, diminishing to the point of disappearance the temperature range where the transformation is considered as being of the bainitic mode. (author) [fr

Low-enriched uranium-molybdenum fuel plate development

International Nuclear Information System (INIS)

Wiencek, T.C.; Prokofiev, I.G.

2000-01-01

To examine the fabricability of low-enriched uranium-molybdenum powders, full-size 450 x 60 x 0.5-mm (17.7 x 2.4 x 0.020-in.) fuel zone test plates loaded to 6 g U/cm 3 were produced. U-10 wt.% Mo powders produced by two methods, centrifugal atomization and grinding, were tested. These powders were supplied at no cost to Argonne National Laboratory by the Korean Atomic Energy Research Institute and Atomic Energy of Canada Limited, respectively. Fuel homogeneity indicated that both of the powders produced acceptable fuel plates. Operator skill during loading of the powder into the compacting die and fuel powder morphology were found to be important when striving to achieve homogeneous fuel distribution. Smaller, 94 x 22 x 0.6-mm (3.7 x 0.87 x 0.025-in.) fuel zone, test plates were fabricated using U-10 wt.% Mo foil disks instead of a conventional powder metallurgy compact. Two fuel plates of this type are currently undergoing irradiation in the RERTR-4 high-density fuel experiment in the Advanced Test Reactor. (author)

Determination of impurities in uranium--niobium (7.5%)--zirconium (2.5%) alloy

Energy Technology Data Exchange (ETDEWEB)

Arragon, Y

1973-10-01

The determination of 11 impurities in uranium--niobium-- zirconium alloys was studied. Elements of which the alloy is composed are considered and information is given on the determination of niobium by niobic acid precipitation. Selective elimination of the three components is discussed. Two liquid-liquid extractions are used. The nioblum is separated by methylisobutylketone in a hydrochloric --hydrofluoric medium and the zirconium and uranium by tributyl phosphate in a nitric medium. The determination of trace elements using electrochemical methods is discussed. Anodic re-dissolution polarography or square wave polarography enabled six elements (cadmium, copper, lead, zinc, bismuth, and thallium) to be determined in a carbonate medium together with aluminium in tetraethylammonium perchlorate, molybdenum in nitric acid, ammonium nitrate, and tungsten in hydrochloric acid with added double sodium and potassium tartrate. Fluorine was determined using ionometric techniques with a specific electrode and carbon was titrated by conductometry after combustion of the sample in an oxygen current. (auth)

Study of the molybdenum retention in alumina

International Nuclear Information System (INIS)

Wilkinson, Maria V.; Mondino, Angel V.; Manzini, Alberto

2002-01-01

The Argentine National Atomic Energy Commission routinely produces 99 Mo by fission of highly enriched uranium contained in targets irradiated in RA-3 reactor. The current process begins with the dissolution of the irradiated target in a basic media, considering the possibility of changing the targets, it could be convenient to dissolve them in acid media. The use of alumina as a first separation step in acid dissolution processes is already known although it is necessary to determine both the type of alumina to be used and the separation conditions. The study of molybdenum retention in alumina was performed at laboratory scale, using Mo-99 as radiotracer. Different kinds of alumina were tried, varying charge solution acidity. Influence of uranium concentration in the loading solution on molybdenum retention was also studied. (author)

Production of molybdenum-99 by heterogeneous and homogeneous uranium fueled reactors

International Nuclear Information System (INIS)

Carlin, G.E.; Bonin, H.W.

2012-01-01

The use of radioisotopes for various procedures in the health care industry has become one of the most important practices in medicine. At the forefront of the medical isotope list is molybdenum-99 and its daughter isotope technetium-99m, which encompass over 80% of radiopharmaceutical procedures. Fission of uranium-235 to produce molybdenum-99 is the most widely used method for producing this radioisotope. The heterogeneous reactor and the aqueous homogeneous reactor are looked at here with emphasis on the use of low enriched uranium as the fuel source. Methods of technetium-99m generation and its medical use are also reviewed. (author)

Production of molybdenum-99 by heterogeneous and homogeneous uranium fueled reactors

Energy Technology Data Exchange (ETDEWEB)

Carlin, G.E.; Bonin, H.W., E-mail: george.carlin@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

2012-07-01

The use of radioisotopes for various procedures in the health care industry has become one of the most important practices in medicine. At the forefront of the medical isotope list is molybdenum-99 and its daughter isotope technetium-99m, which encompass over 80% of radiopharmaceutical procedures. Fission of uranium-235 to produce molybdenum-99 is the most widely used method for producing this radioisotope. The heterogeneous reactor and the aqueous homogeneous reactor are looked at here with emphasis on the use of low enriched uranium as the fuel source. Methods of technetium-99m generation and its medical use are also reviewed. (author)

ELECTROCHEMICAL STUDIES OF URANIUM METAL CORROSION MECHANISM AND KINETICS IN WATER

International Nuclear Information System (INIS)

Boudanova, Natalya; Maslennikov, Alexander; Peretroukhine, Vladimir F.; Delegard, Calvin H.

2006-01-01

During long-term underwater storage of low burn-up uranium metal fuel, a corrosion product sludge forms containing uranium metal grains, uranium dioxide, uranates and, in some cases, uranium peroxide. Literature data on the corrosion of non-irradiated uranium metal and its alloys do not allow unequivocal prediction of the paragenesis of irradiated uranium in water. The goal of the present work conducted under the program 'CORROSION OF IRRADIATED URANIUM ALLOYS FUEL IN WATER' is to study the corrosion of uranium and uranium alloys and the paragenesis of the corrosion products during long-term underwater storage of uranium alloy fuel irradiated at the Hanford Site. The elucidation of the physico-chemical nature of the corrosion of irradiated uranium alloys in comparison with non-irradiated uranium metal and its alloys is one of the most important aspects of this work. Electrochemical methods are being used to study uranium metal corrosion mechanism and kinetics. The present part of work aims to examine and revise, where appropriate, the understanding of uranium metal corrosion mechanism and kinetics in water

Evolution of microstructure of U-Mo alloys in as cast and sintered forms

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Kamath, H.S.; Dey, G.K.

2009-01-01

Over the years U 3 Si 2 compound dispersed in aluminium matrix has been successfully used as potential Low Enriched Uranium (LEU 235 ) base dispersion fuel in new research and test reactors and also for converting High Enriched Uranium (HEU > 85% U 235 ) cores to LEU in most of the existing research and test reactors. The maximum density achievable with U 3 Si 2 -AI dispersion fuel is around 4.8 g U cm -3 . To achieve a uranium density of 8.0 to 9.0 g U cm -3 in dispersion fuel with aluminium as matrix material, it is required to use γ-stabilized uranium metal powders. At Metallic Fuels Division, R and D efforts are on to develop these high density uranium alloys. Molybdenum plays a crucial role in metastabilising the γ-phase of uranium at room temperature which is very much evident when we see the microstructures of different U-Mo alloys with varying molybdenum concentration as solute atom. The paper describes the role of molybdenum in imparting metastability in U-Mo alloys from their microstructures in as cast and sintered forms. The paper also covers the role of tailored microstructure in U-Mo alloy for the purpose of hydriding and dehydriding treatment to generate alloy powders. (author)

A study of phase transformations processes in 0,5 to 4% mo uranium-molybdenum alloys; Etude des processus des transformations dans les alliages uranium-molybdene de teneur 0,5 a 4% en poids de molybdene

Energy Technology Data Exchange (ETDEWEB)

Lehmann, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1959-06-15

Isothermal and continuous cooling transformations process have been established on uranium-molybdenum alloys containing 0,5 to 4 w% Mo. Transformations process of the {beta} and {gamma} solid solutions are described. These processes depend upon molybdenum concentration. Out of the {beta} solid solution phase appears an eutectoid decomposition of {beta} to ({alpha} + {gamma}) or the formation of a martensitic phase {alpha}''. The {gamma} solid solution shows a decomposition of {gamma} to ({alpha} + {gamma}) or ({alpha} + {gamma}'), or a formation of martensitic phases a' or a'{sub b}. The U-Mo equilibrium diagram is discussed, particularly in low concentrations zones. Limits between domains ({alpha} + {gamma}) and ({beta} + {gamma}), ({beta} + {gamma}) and {gamma}, ({beta} + {gamma}) and {beta}, have been determined. (author) [French] Les processus des transformations isothermes, et au cours de refroidissements continus ont ete etablis sur les alliages uranium-molybdene de 0,5 a 4 % en poids de Mo. Ceci a permis de mettre en evidence les processus des transformations de solutions solides {beta} et {gamma}, differents suivant la teneur en molybdene de l'alliage. Dans le premier cas il y a decomposition eutectoide de {beta} en ({alpha} + {gamma}) ou formations d'une phase martensitique {alpha}''. Dans le second cas il y a decomposition de {gamma} soit en ({alpha} + {gamma}) soit en ({alpha} + {gamma}') suivant la temperature, ou bien formation des phases martensitiques {alpha}' ou {alpha}'{sub b}. Le diagramme d'equilibre, uranium-molybdene est sujet a de nombreuses controverses, en particulier dans la zone des faibles concentrations. Les limites entre les domaines ({alpha} + {gamma}) et ({beta} + {gamma}), ({beta} + {gamma}) et {gamma}, ({beta} + {gamma}) et {beta}, ont ete determinees. (auteur)

Microstructural characteristics of DU-xMo alloys with x = 7-12 wt%

International Nuclear Information System (INIS)

Burkes, Douglas E.; Hartmann, Thomas; Prabhakaran, Ramprashad; Jue, J.-F.

2009-01-01

Microstructural, phase, and impurity analyses of six depleted uranium-molybdenum alloys were obtained using optical metallography, X-ray diffraction, and carbon/nitrogen/oxygen determination. Uranium-molybdenum alloy foils are currently under investigation for the conversion of high-power research reactors using high-enriched uranium fuel to accommodate the use of low-enriched uranium fuel. Understanding basic microstructural behavior of these foils is an important consideration in determining the impact of fabrication processes and in anticipating performance of the foils in a reactor. Average grain diameter decreased with increasing molybdenum content. Lattice parameter decreased with increasing molybdenum content, and no significant degree of phase decomposition or crystallographic ordering was caused by processing and post-processing conditions employed in this study. Impurity concentration, specifically carbon, inhibited the degree of microstructural recrystallization but did not appear to impact other microstructural traits, such as γ-phase retention or lattice parameter.

DAMAGE IN MOLYBDENUM ASSOCIATED WITH NEUTRON IRRADIATION AND SUBSEQUENT POST-IRRADIATION ANNEALING

Energy Technology Data Exchange (ETDEWEB)

Mastel, B.

1963-07-23

Molybdemum containing carbon was studied in an attempt to establish the combined effect of impurity content and neutron irradiation on the properties and structure of specific metals. Molybdenum foils were punched into discs and heat treated in vacuum. They were then slow-cooled and irradiated. After irradiation and subsequent decay of radioactivity to a low level the foils were subjected to x-ray diffraction measurements. Cold-worked foils with less than 10 ppm carbon showed no change in microstructure due to irradiation. Molybdenum foils that were annealed prior to irradiation showed spot defects. In foils containing up to 500 ppm carbon, it was concluded that the small loops present after irradiation are due to the clustering of point defects at interstitial carbon atoms, followed by collapse to form a dislocation loop. The amount of lattice expansion after irradiation was strongly dependent on impurity content. Neutron irradiation was found to reduce the number of active slip systems. (M.C.G.)

Fission gas behaviour and interdiffusion layer growth in in-pile and out-of-pile irradiated U-Mo/Al nuclear fuels

International Nuclear Information System (INIS)

Zweifel, Tobias

2014-01-01

Worldwide, research and test reactors are to convert their fuel from highly towards lower enriched uranium, among them the FRM II. One prospective fuel is an alloy of uranium and molybdenum (abbr. U-Mo). Test irradiations showed an insufficient irradiation behavior of this new fuel due to the growth of an interdiffusion layer (abbr. IDL) between the U-Mo fuel and the surrounding Al matrix. Furthermore, this layer accumulates fission gases. In this work, heavy ion irradiated U-Mo/Al layer systems were studied and compared to in-reactor irradiated fuel to study the fission gas dynamics. It is demonstrated that the gas behavior is identical for both in-reactor and out-of-reactor approaches.

Deposit of molybdenum associated with uranium in Pena Blanca, Mexico

International Nuclear Information System (INIS)

Reyes-Cortes, M.

1985-01-01

The uranium-molybdenum deposits are in the Sierra Pena Blanca, 45 km north of the city of Chihuahua. The largest amounts of uranium-molybdenum ore are found in the area of Las Margaritas-Puerto III. The ratio of molybdenum mineralization to uranium is 2:1 in this area and the deposits are distributed at depths of 55-100 m in ignimbritic rocks of the so called Escuadra Formation. This volcanic unit consists of an altered crystalline-lithic ash-flow tuff of Oligocene age. The molybdenum mineral occurs as powellite (CaMoO 4 ) and is found predominantly in two size ranges: phenocrysts 0.1-20 mm in diameter are abundant in the upper part of the deposit, while a material which varies between cryptocrystalline and amorphous predominates in the lower part. This latter material can easily be identified inside the mine by its strong orange fluorescence; it is also easy to recover by leaching. In contrast, the metallurgical process of recovery by leaching of the phenocrystalline portion of the powellite has so far presented problems. Powellite is generally found in association with carnotite, margaritasite and uranophane, and its mineralization consists of disseminated lumps, druses, crustifications and veins; frequently, it partially replaces the phenocrysts of argillized feldspars of the Escuadra Formation. Fractured and brecciated zones with intense oxidation of jarosite, haematite, limonite and goethite sometimes show high U-Mo concentrations; on other occasions the concentration is found with alunite at the contact between the ignimbrite and the layers of argillized vitrophyre. The mineralizations of fluorite, pyrite, jarosite, alunite and opal are indicative of hydrothermal deposition, possibly at low temperature with supergene or geothermal alterations. (author)

Analysis of prospecting effect of polonium survey and geoelectric survey extracted uranium and molybdenum in the south of Shengyuan volcanic basin

International Nuclear Information System (INIS)

Jin Hehai

2007-01-01

Polonium survey and geoelectric survey extracted uranium and molybdenum show that compound anomaly with sharp anomaly peak of the curve of polonium-210, uranium, molybdenum appears along many survey lines in Bakou area, Shengyuan volcanic basin, which may reflect the enrichment of uranium and molybdenum in rock formation and soil layer. By contrasting the anomaly curve to that above the buried uranium deposit, it is recognized that compound anomaly is closely related to the uranium mineralization condition in the area and some favourable sites for uranium metallogeny have been predicated. (authors)

Study of the recrystallisation of irradiated uranium; Etude sur l'uranium irradie

Energy Technology Data Exchange (ETDEWEB)

Bloch, J; Mustelier, J P; Bussy, P; Blin, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

1- Study of the recrystallisation of irradiated uranium. The recrystallisation of uranium irradiated to a burnup level of 220 MWj/t, at a temperature of the order of 350 deg. C, has been investigated. The observations were made chiefly by means of micrography an hardness measurements. If the irradiated metal is compared with a cold-drawn metal showing the same shearing of the twinned crystals, and therefore the same rate of plastic deformation, as the irradiated metal, it is noted that the restoring of the irradiated metal takes place at a considerably higher temperature than that of the cold-drawn metal. Pre-crystallisation is very much delayed. Only, a passage of the {alpha}-{beta} transformation point quickly wipes out irradiation effect. 2- Hardening of uranium by irradiation. Using hardness measurements we have studied more especially the effect of very weak irradiations on uranium (integrated flux < 10{sup 16} nvt). The hardness does not increase linearly with the flux, but a period of incubation is observed probably representing the time necessary for saturation of the dislocations. (author)Fren. [French] 1- Etude de la recristallisation de l'uranium irradie. On a etudie la recristallisation d'uranium irradie jusqu'a un taux de combustion de 220 MWj/t a une temperature de l'ordre de 350 deg. C. Les observations ont ete faites principalement a l'aide de la micrographie et de la durete. Si l'on compare le metal irradie avec un metal ecroui presentant le meme cisaillement des macles, donc le meme taux de deformation plastique que le metal irradie, on constate que la restauration du metal irradie se fait a une temperature notablement superieure a celle du metal ecroui. La recristallisation est tres retardee. Seul, un passage du point de transformation {alpha}-{beta} efface rapidement l'effet de l'irradiation. 2- Durcissement de l'uranium par irradiation. Nous avons, a l'aide de la durete, etudie plus particulierement l'effet de tres faibles irrtions sur l'uranium

Paraelasticity in electron-irradiated molybdenum

International Nuclear Information System (INIS)

Beuneu, Brigitte; Quere, Yves.

1981-11-01

The relaxation of a radiation-induced point defect-most probably the rotation of a dumbell-is observed during isothermal anneals of irradiated molybdenum by resistivity measurements. The recovery of close pairs is not affected, in first analysis, by the presence of a uniaxial stress

Investigation into electrochemical behavior of molybdenum VM-1 alloy at high current density

Energy Technology Data Exchange (ETDEWEB)

Tatarinova, O M; Amirkhanova, N A; Akhmadiev, A G

1975-01-01

The effect of the composition and concentration of electrolyte on the workability of the molybdenum VM-1 alloy has been studied and a number of anions has been determined relative to their activation capacity. The best workability of the alloy is achieved in a 15% NaOH solution and a composite electrolyte 15% NaNO/sub 3/+5%NaOH. It is shown that in polarization of the VM-1 alloy both in alkali- and salt solutions a film of oxides of different valence molybdenum is formed: Mo/sub 2/O/sub 3/, Mo/sub 4/O/sub 11/, Mo/sub 9/O/sub 26/, MoO/sub 3/, but molybdenum gets dissolved only in a hexavalent form, its content in a solution being in conformity with the polarizing current densities. Using a temperature-kinetic technique it has been found that the concentrational polarization is the limiting stage in the reaction of molybdenum and VM-1 alloy anodic dissolution in 15% NaNO/sub 3/ solution and in the composite electrolyte 15%NaNO/sub 3/+5%NaOH.

Irradiation tests of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, A.; Nasution, H.; Lufti-Amin, D.; Gogo, A.

1996-01-01

Most of the world's supply of 99 mTc for medical purposes is currently produced from the decay of 99 Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO 2 used in current target designs will allow equivalent 99 Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers. (author)

Irradiation tests of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, A.; Nasution, H.; Lufti-Amin, D.; Gogo, A.

1996-01-01

Most of the world's supply of 99m Tc for medical purposes is currently produced form the decay of 99 Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO 2 used in current target designs will allow equivalent 99 Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers

X-ray target with substrate of molybdenum alloy

International Nuclear Information System (INIS)

Hirsch, H.H.

1980-01-01

Rotary targets for x-ray tubes are provided comprising a molybdenum base body alloyed with a stabilizing proportion of iron, silicon, cobalt, tantalum, niobium, hafnium, stable metal oxide, or a mixture of the preceding

Oxidation of uranium and uranium alloys

International Nuclear Information System (INIS)

Orman, S.

1976-01-01

The corrosion behaviour of uranium in oxygen, water and water + oxygen mixtures is compared and contrasted. A considerable amount of work, much of it conflicting, has been published on the U + H 2 O and U + H 2 O + O 2 systems. An attempt has been made to summarise this data and to explain the reasons for the lack of agreement between the experimental results. The evidence for the mechanism involving OH - ion diffusion as the reacting entity in both the U + H 2 O and U + O 2 + H 2 O reactions is advanced. The more limited corrosion data on some lean uranium alloys and on some higher addition alloys referred to as stainless materials is summarised together with some previously unreported results obtained with these materials at AWRE. The data indicates that in the absence of oxygen the lean alloys behave in a similar manner to uranium and evolve hydrogen in approximately theoretical quantities. But the stainless alloys absorb most of the product hydrogen and assessments of reactivity based on hydrogen evolution would be very inaccurate. The direction that future corrosion work on these materials should take is recommended

The reprocessing of irradiated fuels improvement and extension of the solvent extraction process

International Nuclear Information System (INIS)

Faugeras, P.; Chesne, A.

1964-01-01

Improvements made in the conventional tri-butylphosphate process are described, in particular. the concentration and the purification of plutonium by one extraction cycle using tri-butyl-phosphate with reflux; and the use of an apparatus working continuously for precipitating plutonium oxalate, for calcining the oxalate, and for fluorinating the oxide. The modifications proposed for the treatment of irradiated uranium - molybdenum alloys are described, in particular, the dissolution of the fuel, and the concentration of the fission product solutions. The solvent extraction treatment is used also for the plutonium fuels utilized for the fast breeder reactor (Rapsodie) An outline of the process is presented and discussed, as well as the first experimental results and the plans for a pilot plant having a capacity of 1 kg/day. The possible use of tn-lauryl-amine in the plutonium purification cycle is now under consideration for the processing plant at La Hague. The flowsheet for this process and its performance are presented. The possibility of vitrification is considered for the final treatment of the concentrated radioactive wastes from the Marcoule (irradiated uranium) and La Hague (irradiated uranium-molybdenum) Centers. Three possible processes are described and discussed, as well as the results obtained from the operation of the corresponding experimental units using tracers. (authors) [fr

Weldability of molybdenum and its alloy sheet, 1

International Nuclear Information System (INIS)

Matsuda, Fukuhisa; Ushio, Masao; Nakata, Kazuhiro; Edo, Yoshiaki

1979-01-01

Basic weldability of electron-beam melted pure molybdenum has been examined in electron-beam welding in high vacuum and GTA welding in pure and air mixed argon atmospheres by paying attention to weld defects such as hot cracking and porosity in weld metal and also mechanical properties of welded joint in comparison with conventional TZM alloys. The main conclusions obtained were as follows; (1) The weld metals of electron-beam melted pure molybdenum with electron-beam and GTA weldings in pure and air mixed argon atmosphere up to about 1% were almost porosity free. However, large amount of oxygen content of 200 ppm in powder-metallurgy TZM alloy made very porous weld bead in electron-beam welding in high vacuum. Therefore, oxygen content in base metal should be lowered to the minimum, that is, less than 10 ppm, especially in electron-beam welding in high vacuum. (2) Hot cracking occurred in the weld metal of GTA welding when air content in argon atmosphere exceeded about 0.6% for electron-beam melted pure molybdenum and powder metallurgy TZM alloy. In less than 0.26% air, no hot cracking were observed in this experiment. Moreover, in electron-beam welding, no hot cracking was observed in weld metals for both materials. In order to prevent the formation of hot cracking, the purity of welding atmosphere should be kept as high as possible. (3) Joint efficiency of the welded joint of electron-beam melted pure molybdenum with electron-beam welding was 50 to 60% to base metal at room temperature and 500 0 C and almost 100% at 1000 0 C. Those of GTA welds in pure and 0.13% air mixed argon atmospheres were fairly lower than those in electron-beam welding for each testing temperature. (author)

Neutron irradiation damage of a stress relieved TZM alloy

International Nuclear Information System (INIS)

Abe, K.; Masuyama, T.; Satou, M.; Hamilton, M.L.

1992-01-01

The objective of this work is to study defect microstructures and irradiation hardening in a stress relieved TZM alloy after irradiation in the Fast Flux Test Facility (FFTF) using the Materials Open Test Assembly (MOTA). Disk specimens of the molybdenum alloy TZM that had been stress relieved at 1199 K (929 C) for 0.9 ks (15 min.) were irradiated in the FFTF/MOTA 1F at 679, 793 and 873 K (406, 520, and 600 C) to a fast fluence of ∼9.6 x 10 22 n/cm 2 . Microstructures were observed in a transmission electron microscope (TEM). Dislocation structures consisted of isolated loops, aggregated loops (rafts) and elongated dislocations. The size of the loops increased with the irradiation temperature. Void swelling was about 1 and 2% at 793 and 873 K (520 and 600 C), respectively. A void lattice was developed in the body centered cubic (bcc) structure with a spacing of 26 - 28 nm. The fine grain size (0.5 - 2 μm) was retained following high temperature irradiation, indicating that the stress relief heat treatment may extend the material's resistance to radiation damage up to high fluence levels. Microhardness measurements indicated that irradiation hardening increased with irradiation temperature. The relationship between the microstructure and the observed hardening was determined

Thermal cycling behaviour and thermal stability of uranium-molybdenum alloys of low molybdenum content; Comportement au cyclage thermique et stabilite thermique des alliaces uranium-molybdene de faibles teneurs en molybdene

Energy Technology Data Exchange (ETDEWEB)

Decours, J; Fabrique, B; Peault, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1963-07-01

We have studied the behaviour during thermal cycling of as-cast U-Mo alloys whose molybdenum content varies from 0.5 to 3 per cent; results are given concerning grain stability during extended heat treatments and the effect of treatments combining protracted heating with thermal cycling. The thermal cycling treatments were carried out at 550, 575, 600 and 625 deg C for 1000 cycles; the protracted heating experiments were done at 550, 575, 600 and 625 deg C for 2000 hours (4000 hrs at 625 deg C). The 0.5 per cent alloy resists much better to the thermal cycling than does the non-alloyed uranium. This resistance is, however, much lower than that of alloys containing over l per cent, even at 550 deg C it improves after a heat treatment for grain-refining. Alloys of over 1.1 per cent have a very good resistance to a cycling treatment even at 625 deg C, and this behaviour improves with increasing concentrations up to 3 per cent. An increase in the temperature up to the {gamma}-phase has few disadvantages provided that it is followed by rapid cooling (50 to 100 deg C/min). The {alpha} grain is fine, the {gamma}-phase is of the modular form, and the behaviour during a thermal cycling treatment is satisfactory. If this cooling is slow (15 deg /hr) the {alpha}-grain is coarse and cycling treatment behaviour is identical to that of the 0.5 per cent alloy. The protracted heat treatments showed that the {alpha}-grain exhibits satisfactory stability after 2000 hours at 575, 600 and 625 deg C, and after 4000 hours at 625 deg C. A heat cycling treatment carried out after these tests affects only very little the behaviour of these alloys during cycling. (authors) [French] Nous avons etudie le comportement au cyclage thermique des alliages U-Mo, brut de coulee, dont la teneur varie de 0,5 a 3 pour cent de molybdene, les resultats de stabilite du grain au cours de traitements thermiques de longue duree, ainsi que ceux des traitements combines de longue duree et de cyclage. Les

Determination of ultratrace amounts of uranium and thorium in aluminium and aluminium alloys by electrothermal vaporization/ICP-MS

International Nuclear Information System (INIS)

Nakamura, Yasushi; Kobayashi, Yoshio; Kakurai, Yousuke

1993-01-01

A method has been developed for determining the 0.01 ng g -1 level of uranium and thorium in aluminium and aluminium alloys by electrothermal vaporization (ETV)/ICP-MS. This method was found to be significantly interfered with any matrices or other elements contained. An ion-exchange technique was therefore applied to separate uranium and thorium from aluminium and other elements. It was known that uranium are adsorbed on an anion-exchange resin and thorium are adsorbed on cation-exchange resin. However, aluminium and copper were eluted with 6 M hydrochloric acid. Dissolve the sample with hydrochloric acid containing copper which was added for analysis of pure aluminium, and oxidize with hydrogen peroxide. Concentration of hydrochloric acid in the solution was adjusted to 6 M, and then passed the solution through the mixed ion-exchange resin column. After the uranium and thorium were eluted with 1 M hydrofluoric acid-0.1 M hydrochloric acid, the solution was evaporated to dryness. It was then dissolved with 1 M hydrochloric acid. Uranium and thorium were analyzed by ETV/ICP-MS using tungsten and molybdenum boats, respectively, since the tungsten boat contained high-level thorium and the molybdenum boat contained uranium. The determination limit of uranium and thorium were 0.003 and 0.005 ng g -1 , respectively. (author)

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

Science.gov (United States)

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.

Ion exchange resin fouling of molybdenum in recovery uranium processess

International Nuclear Information System (INIS)

Zhang Guowei; Zhao Guirong

1990-09-01

The relationship between anion exchange resin fouling and molybdic acid polymerization was studied. By using potentiometer titration and laser-Raman spectroscopy the relationship of molybdic acid polymerization and the pH value of solution or the molybdenum concentration was determined. It was shown that as the concentration of initial molybdenum in solution decreases from 0.2 mol/L to 0.5 mmol/L, the pH value of starting polymerization decreased from 6.5 to 4.5. The experimental results show that the fouling of 201 x 7 resin in the acidic solution is mainly caused by the adsorbing of Mo 3 O 26 4- ion and occupying the exchange radical site of the resin. Under the leaching conditions the molybdenum and phosphate existing in the leaching liquor can form 12-molybdo-phosphate ion. It also leads to resin fouling. The molybdenum on the fouled resin can synergically be desorbed by mixed desorbents containing ammonium hydroxide and ammonium sulfate. The desorbed resin can be used for uranium adsorption and the desorbed molybdenum can be recovered by ion exchange method

The use of slightly alloyed uranium as fuel: its influence on the dissolution and other stages of treatment

International Nuclear Information System (INIS)

Faugeras, P.; Leroy, P.; Lheureux, C.

1959-01-01

This report deals chiefly with the treatment of binary alloys (UAI, UMo, UZr, UCr, USi) with a low concentration of the additional element (≤2 per cent). The investigation was pursued with a view to the continued utilisation, with a minimum of modification, of the existing plants for treatment of non-alloyed irradiated uranium. In the first part, the usual process for the treatment of irradiated uranium by solvent extraction is briefly recalled. The second part is devoted to a study of the selective dissolution of the canning around certain of these alloys. The third part gives the behaviour of these different alloys at various phases of the usual treatment: a) dissolution; b) extractions; c) final treatment of fission products; d) final purification of plutonium. To conclude, possible alloys are classed as a function of their repercussions on the normal treatment. (author) [fr

Effect of low-temperature thermomechanical treatment on mechanical properties of low-alloying molybdenum alloys with carbide hardening

International Nuclear Information System (INIS)

Bernshtejn, L.M.; Zakharov, A.M.; Veller, M.V.

1978-01-01

Presented are results of testing low-temperature thermomechanical treatment of low-alloying molybdenum alloys, including quenching from 2100 deg C, 40% deformation by hydroextrusion and aging at the temperature of 1200-1400 deg C. Tensile tests at room temperature with the following processing of results have shown that low-temperature thermomechanical treatment of low-alloying molybdenum alloys of Mo-Zr-C and Mo-Zr-Nb-C systems leads to a significant increase in low-temperature mechanical properties (strength properties - by 30-35%, ductility - by 30-40%) as compared with conventional heat treatment (aging after quenching). The treatment proposed increases resistance to small, as well as large plastic deformations, and leads to a simultaneous rise of strength and plastic properties at all stages of tensile test. Alloying of the Mo-Zr-C system with niobium increases both strength and plastic characteristics as compared with alloys without niobium when testing samples, subjected to low temperature thermomechanical treatment and conventional heat treatment at room temperature

Process for continuous production of metallic uranium and uranium alloys

Science.gov (United States)

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

1995-01-01

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

Process for continuous production of metallic uranium and uranium alloys

Science.gov (United States)

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.

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

Energy Technology Data Exchange (ETDEWEB)

Okuniewski, Maria A. [Purdue Univ., West Lafayette, IN (United States); Ganapathy, Varsha [Purdue Univ., West Lafayette, IN (United States); Hamilton, Brenden [Purdue Univ., West Lafayette, IN (United States); Cassutt, Paul [Purdue Univ., West Lafayette, IN (United States); Zhang, Fan [Purdue Univ., West Lafayette, IN (United States); Velaquez, Daniel [Illinois Inst. of Technology, Chicago, IL (United States); Seibert, Rachel [Illinois Inst. of Technology, Chicago, IL (United States); Terry, Jeff [Illinois Inst. of Technology, Chicago, IL (United States); Sprouster, David [Brookhaven National Lab. (BNL), Upton, NY (United States); Ecker, Lynne [Brookhaven National Lab. (BNL), Upton, NY (United States); Elbakhshwan, Mohamed [Brookhaven National Lab. (BNL), Upton, NY (United States)

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 to 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.

Properties of low content uranium-molybdenum alloys which may be used as nuclear fuels; Proprietes des alliages uranium-molybdene de faibles teneurs utilisables comme materiaux combustibles

Energy Technology Data Exchange (ETDEWEB)

Lehmann, J; Decours, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

Metallurgical properties are given in this report of uranium-molybdenum alloys containing 0,5 to 3 per cent of molybdenum. Since some of these alloys are used in EDF power reactors are given: briefly the operating conditions imposed on nuclear fuels: maximum temperature, temperature gradient and external pressure. In the first part are considered the structural properties of the alloys correlation with the phase transformation kinetics; a description is given of the effects of certain physico-metallurgical factors on the morphology and the crystalline structure of the materials: - solidification conditions and the heredity of the {gamma} structure, - cooling rate at the transformation points, - whether or not the intermediate {gamma} {yields} {beta} transformation is suppressed In the second part we show how a knowledge of the phase transformation processes has made it possible to define the optimum preparation conditions for these materials in the form of fuel tubes intended for the EDF reactors: casting conditions, controlled cooling treatments, weldability. In the third part we study the thermal, stability during the long duration high temperature treatments and the cycles in the two zones of the diagram {alpha} + {gamma}; {beta} + {gamma} the effects of the morphology (in particular the two types of {alpha} pseudo-grains observed) and of the cooling rate during the transformation point transitions are described. In the fourth part are discussed the mechanical properties: resistance to a tractive force, resistance to creep, resilience. These properties can also be affected by the {gamma} structure heredity and by the cooling rate to which the alloy has been subjected. In conclusion we discuss the reasons which led to the choice of some of these alloys for the first EDF reactors in particular the advantages of their high creep resistance between 450 and 600 deg C for use in the form of tubes subjected to an external pressure. (authors) [French] Dans ce rapport

Friction and corrosion resistance of sputter deposited supersaturated metastable aluminium-molybdenum alloys

Energy Technology Data Exchange (ETDEWEB)

Abu-Zeid, O.A. [Univ. of the United Arab Emirates, Al-Ain (United Arab Emirates). Dept. of Mech. Eng.; Bates, R.I. [Design, Mfg. and Marketing Research Inst., Univ. of Salford (United Kingdom)

1996-12-15

Two closed field unbalanced magnetrons with targets of aluminium and molybdenum have been used for the co-deposition of aluminium-molybdenum coatings with different compositions. A pin on disk machine and a computer controlled potentiostat have been used to evaluate respectively, the tribological and corrosion properties of the deposited alloys. Results have shown that introducing molybdenum into aluminium coatings improves their poor tribological properties. Aluminium-molybdenum coatings with different compositions have shown low wear behaviour and for coatings with high molybdenum contents (> 80%) friction coefficients against steel, as low as 0.18 have been obtained. The addition of molybdenum into aluminium coatings has reduced their corrosion tendency and corrosion current density in a marine environment. (orig.)

Solvent-extraction and purification of uranium(VI) and molybdenum(VI) by tertiary amines from acid leach solutions

International Nuclear Information System (INIS)

La Gamma, Ana M.G.; Becquart, Elena T.; Chocron, Mauricio

2008-01-01

Considering international interest in the yellow-cake price, Argentina is seeking to exploit new uranium ore bodies and processing plants. A study of similar plants would suggest that solvent- extraction with Alamine 336 is considered the best method for the purification and concentration of uranium present in leaching solutions. In order to study the purification of these leach liquors, solvent-extraction tests under different conditions were performed with simulated solutions which containing molybdenum and molybdenum-uranium mixtures. Preliminary extraction tests carried out on mill acid-leaching liquors are also presented. (authors)

Influence of plastic deformation on nitriding of a molybdenum-hafnium alloy

International Nuclear Information System (INIS)

Lakhtin, Yu.M.; Kogan, Ya.D.; Shashkov, D.P.; Likhacheva, T.E.

1982-01-01

The influence of a preliminary plastic strain on the structure and properties of molybdenum alloy with 0.2 wt.% Hf upon nitriding in the ammonia medium at 900-1200 deg C during 1-6 h is investigated. The study of microhardness distribution across the nitrided layer thickness has shown that with increase of the degree of preliminary plastic strain up to 50 % the nitrided layer hardness decreases and with further reduction growth up to 90 % - increases. Nitriding sharply (hundred times) increases wear resistance of molybdenum alloy with hafnium addition. At the reduction degree 25 % the wear resistance is less than at other values of percentage reduction in area owing to the minimum thickness of the nitride zone. The alloy strained before nitriding by 25 % has shown the best results during heat resistance testing

Thermodynamic properties of uranium in gallium–aluminium based alloys

International Nuclear Information System (INIS)

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

2015-01-01

Activity, activity coefficients and solubility of uranium was determined in gallium-aluminium alloys containing 1.6 (eutectic), 5 and 20 wt.% aluminium. Additionally, activity of uranium was determined in aluminium and Ga–Al alloys containing 0.014–20 wt.% Al. Experiments were performed up to 1073 K. Intermetallic compounds formed in the alloys were characterized by X-ray diffraction. Partial and excess thermodynamic functions of U in the studied alloys were calculated. - Highlights: • Thermodynamics of uranium is determined in Ga–Al alloys of various compositions. • Uranium in the mixed alloys interacts with both components, Ga and Al. • Interaction of U with Al increases with decreasing temperature. • Activity and solubility of uranium depend on Al content in Ga–Al alloys.

Thermodynamic properties of uranium in gallium–aluminium based alloys

Energy Technology Data Exchange (ETDEWEB)

Volkovich, V.A., E-mail: v.a.volkovich@urfu.ru [Department of Rare Metals and Nanomaterials, Institute of Physics and Technology, Ural Federal University, Ekaterinburg, 620002 (Russian Federation); Maltsev, D.S.; Yamshchikov, L.F. [Department of Rare Metals and Nanomaterials, Institute of Physics and Technology, Ural Federal University, Ekaterinburg, 620002 (Russian Federation); Chukin, A.V. [Department of Theoretical Physics and Applied Mathematics, Institute of Physics and Technology, Ural Federal University, Ekaterinburg, 620002 (Russian Federation); Smolenski, V.V.; Novoselova, A.V. [Institute of High-Temperature Electrochemistry UD RAS, Ekaterinburg, 620137 (Russian Federation); Osipenko, A.G. [JSC “State Scientific Centre - Research Institute of Atomic Reactors”, Dimitrovgrad, 433510 (Russian Federation)

2015-10-15

Activity, activity coefficients and solubility of uranium was determined in gallium-aluminium alloys containing 1.6 (eutectic), 5 and 20 wt.% aluminium. Additionally, activity of uranium was determined in aluminium and Ga–Al alloys containing 0.014–20 wt.% Al. Experiments were performed up to 1073 K. Intermetallic compounds formed in the alloys were characterized by X-ray diffraction. Partial and excess thermodynamic functions of U in the studied alloys were calculated. - Highlights: • Thermodynamics of uranium is determined in Ga–Al alloys of various compositions. • Uranium in the mixed alloys interacts with both components, Ga and Al. • Interaction of U with Al increases with decreasing temperature. • Activity and solubility of uranium depend on Al content in Ga–Al alloys.

Method of fabricating a uranium-bearing foil

Science.gov (United States)

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

2012-04-24

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

Irradiation testing of high density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

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

1997-10-01

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

Irradiated uranium reprocessing

International Nuclear Information System (INIS)

Gal, I.

1961-12-01

Task concerned with reprocessing of irradiated uranium covered the following activities: implementing the method and constructing the cell for uranium dissolving; implementing the procedure for extraction of uranium, plutonium and fission products from radioactive uranium solutions; studying the possibilities for using inorganic ion exchangers and adsorbers for separation of U, Pu and fission products

Effect of molybdenum on structure, microstructure and mechanical properties of biomedical Ti-20Zr-Mo alloys

Energy Technology Data Exchange (ETDEWEB)

Kuroda, Pedro Akira Bazaglia [UNESP - Univ Estadual Paulista, Laboratório de Anelasticidade e Biomateriais, 17.033-360, Bauru, SP (Brazil); IBTN/Br – Institute of Biomaterials, Tribocorrosion and Nanomedicine, Brazilian Branch, 17.033-360 Bauru, SP (Brazil); Buzalaf, Marília Afonso Rabelo [USP – Universidade de São Paulo, Departamento de Ciências Biológicas, 17.012-901, Bauru, SP (Brazil); Grandini, Carlos Roberto, E-mail: betog@fc.unesp.br [UNESP - Univ Estadual Paulista, Laboratório de Anelasticidade e Biomateriais, 17.033-360, Bauru, SP (Brazil); IBTN/Br – Institute of Biomaterials, Tribocorrosion and Nanomedicine, Brazilian Branch, 17.033-360 Bauru, SP (Brazil)

2016-10-01

Titanium has an allotropic transformation around 883 °C. Below this temperature, the crystalline structure is hexagonal close-packed (α phase), changing to body-centered cubic (β phase). Zirconium has the same allotropic transformation around 862 °C. Molybdenum has body-centered cubic structure, being a strong β-stabilizer for the formation of titanium alloys. In this paper, the effect of substitutional molybdenum was analyzed on the structure, microstructure and selected mechanical properties of Ti-20 Zr-Mo (wt%) alloys to be used in biomedical applications. The samples were prepared by arc-melting and characterized by x-ray diffraction with subsequent refinement by the Rietveld method, optical and scanning electron microscopy. The mechanical properties were analyzed by Vickers microhardness and dynamic elasticity modulus. X-ray measurements and Rietveld analysis revealed the presence of α′ phase without molybdenum, α′ + α″ phases with 2.5 wt% of molybdenum, α″ + β phases with 5 and 7.5 wt% of molybdenum, and only β phase with 10 wt% of molybdenum. These results were corroborated by microscopy results, with a microstructure composed of grains of β phase and lamellae and needles of α′ and α″ phase in intra-grain the region. The hardness of the alloy was higher than the commercially pure titanium, due to the action of zirconium and molybdenum as hardening agents. The samples have a smaller elasticity modulus than the commercially pure titanium. - Highlights: • Ti-20Zr-Mo system alloys were developed. • β-Stabilizer effect of Zr in the presence of another β-stabilizer element • Alloys with low elastic modulus.

Fuel powder production from ductile uranium alloys

International Nuclear Information System (INIS)

Clark, C.R.; Meyer, M.K.

1998-01-01

Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

The effect of cooling rate from the γ-phase on the strain-rate sensitivity of a uranium 2 sup(w)/o molybdenum alloy

International Nuclear Information System (INIS)

Boyd, G.A.C.; Harding, J.

1983-01-01

Tensile tests have been performed at strain rates from 10 -4 to about 2000/s and temperatures from -150 deg C to +250 deg C on a uranium 2 w/o molybdenum alloy which had been aged for 2 hours at 500 deg C after a fast gas cool from the γ-phase at a controlled rate of 40 deg C/minute. The results are compared with those for standard as-extruded material which had received the same aging treatment. Stress-strain curves are presented and the effect of strain rate and temperature on the flow stress, the ultimate tensile stress and the elongation to fracture is determined. A thorough structural characterisation of the specimen materials, using X-ray analysis and scanning and transmission electron microscopy, allows the different mechanical responses to be related to the corresponding microstructural state of the material. Flow stress data at different temperatures and strain rates are analysed in terms of the theory of thermally-activated flow and estimates made of the various activation parameters. (author)

Manufacture and properties of molybdenum-rhenium alloys

International Nuclear Information System (INIS)

Fischer, B.; Freund, D.

2001-01-01

It is necessary to measure strength and creep behavior to guarantee the safe and reliable usage of refractory alloys at extremely high temperatures. In the literature there is very little information available about the properties of Mo-Re alloys at temperatures higher than 1000 C. A special test facility has been designed and built for stress-rupture testing at very high temperatures (up to 3000 C) of refractory metals and alloys in inert atmospheres. - The stress-rupture strength as well as the creep behavior of molybdenum-rhenium alloys with rhenium contents between 41 and 51 wt.% have been determined at temperatures ranging from 1200 to 2000 C, and rupture times of up to 10 hours using this facility. Previous measurements of stress-rupture strength and creep behavior of pure rhenium have been compared with the measurement results of Mo-Re alloys. - The discussion of the values measured is based on metallographic test results and scanning electron microscopy (SEM) images of Mo-Re alloy samples after stress-rupture testing. (orig.) [de

Molybdenum-UO2 cerment irradiation at 1145 K

Science.gov (United States)

Mcdonald, G.

1971-01-01

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

Evaluation of 99Mo/99mTc generator experiment using PZC material and irradiated natural molybdenum

International Nuclear Information System (INIS)

Khongpetch, P.; Chingjit, S.; Dangprasert, M.; Rangsawai, W.; Virawat, N.

2006-01-01

Technetium-99m ( 99m Tc) is the most widely used radioisotope in nuclear medicine, accounting for more than 80% of all diagnostic nuclear medicine procedure. 99m Tc is almost exclusively produced from the decay of its parent molybdenum-99 ( 99 Mo). The present sources of 99 Mo are research reactors by using the (n, γ) nuclear reaction with natural molybdenum, resulting in inexpensive but low specific activity 99 Mo, or by neutron-induced fission of uranium-235, which result in expensive but high specific activity 99 Mo. The technology requirement for processing of 99 Mo from the (n, γ) 'activation method' is rather simple, and is within the reach of most developing countries operating research reactors. In the fission method' the technological and infrastructure requirements are some complex, and possibly can be sustained only by countries with advanced nuclear technology. To overcome these difficulties, Japan Atomic Energy Research Institute (JAERI) and KAKEN company have developed alternative technology for 99 Mo/ 99m Tc generator by using a molybdenum absorbent called Poly Zirconium compound (PZC) and irradiated natural molybdenum. The paper describes experiments for evaluation the performance of PZC as a column packing material for 99 Mo/ 99m Tc generator from (n, γ) 99 Mo. (author)

Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures; Desenvolvimento de um combustivel de alta densidade a base das ligas uranio-molibdenio com alta compatibilidade em altas temperaturas

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Fabio Branco Vaz de

2008-07-01

This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature

Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

Energy Technology Data Exchange (ETDEWEB)

Mushtaq, A. [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)], E-mail: mushtaqa@pinstech.org.pk; Iqbal, Massod; Bokhari, Ishtiaq Hussain; Mahmood, Tariq; Mahmood, Tayyab; Ahmad, Zahoor; Zaman, Qamar [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

2008-02-15

Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% {sup 235}U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required {sup 99}Mo/{sup 99m}Tc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99.

Irradiation testing of high-density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

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

1997-01-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

Ultrasonic measurement of elastic moduli of 17-4 pH stainless steel and uranium -2 molybdenum from -400C to 8000C

International Nuclear Information System (INIS)

Gieske, J.H.

1980-10-01

Young's Modulus, shear modulus, and Poisson's ratio for 17-4 pH stainless steel and uranium -2 molybdenum are calculated from ultrasonic longitudinal and shear velocities determined from -40 0 C to 800 0 C. The ultrasonic velocities were determined at elevated temperatures using a through-transmission buffer rod arrangement. An indium-gallium slurry bond was used as an ultrasonic couplant between Cupernickel 10 alloy buffer rods and the specimen. Microstructural changes and phase transitions in the specimens are evident from the temperature dependence of the ultrasonic data. 10 figures, 3 tables

Interaction between uranium oxide alloyed with Al2O3·SiO2 and pyrocarbon coating during irradiation of micro fuel elements

International Nuclear Information System (INIS)

Chernikov, A.S.; Khromov, Y.F.; Svistunov, D.E.; Chuiko, E.E.

1989-01-01

The thermodynamics of the interaction between uranium oxide and carbon was previously studied in the presence of Al 2 O 3 ·SiO 2 , SiC, and UC 1.86 ; in this case, the quantity of the reacting substances does not have any effect on the attainment of the equilibrium state. Based on the obtained results, it is interesting to study the characteristic features of the interaction between the alloyed UO x cores (kernels) with the PyC-coating under the conditions involving irradiation of the micro fuel elements with thermal neutrons and the formation of solid fission products. The data concerning the characteristics of a micro fuel element (the weight of the core, its composition, etc.) are useful for carrying out a quantitative evaluation of the additives required for fixing the alkali-earth fission products by obtaining stable compounds of aluminosilicates with Ba, Sr, Rb, and Cs at different levels of depletion (burnup) of the oxide fuel. An analysis of the interaction processes in such a complex system as the irradiated alloyed uranium oxide fuel located in a micro fuel element is carried out by comparing the chemical potential of oxygen (RT ln P O 2 ) for the competing constituents of the system

Separation of fission Molybdenum for production of technetium generator

International Nuclear Information System (INIS)

Bayat, L.; Shaham, V.; Davarkha, R.

2002-01-01

There are two basically different methods for Mo-99 productions: Activation of Mo-99 contained at about 24% in natural isotopic mixtures. Mo-98 enriched targets are irradiated in high-flux reactors in order to achieve the highest possible specific activity of the product. Idolisation of fission molybdenum from irradiated nuclear fuel targets which have undergone short-term cooling. Maximum fission yield can be attained by irradiation of uranium-235 with the highest possible enrichment. On account of its approximately 1000 times higher specific activity. Fission molybdenum has almost replaced worldwide the product fabricated by activation. However, fission molybdenum-99 production has as its prerequisite a suitably advanced technology by which the production process taking place under high activity conditions can be controlled. An integral part of the process consist in the retention of the fission gases the recycling of non-consumed fuel and the treatment of the waste streams arising. This publication will deal with the individual steps in the process

Extraction and selective stripping of uranium and molybdenum in sulfate solution using amines

International Nuclear Information System (INIS)

Sialino, E.; Mignot, C.; Michel, P.; Vial, J.

1977-01-01

The uranium solutions issued from leaching of AKOUTA ores and containing lot of molybdenum are purified using solvent extraction. During the first test run precipitation of complexes such as amine phosphomolybdate was observed. It was pointed out that the precipitation could be prevented if the molybdenum is oxidized in the feed prior to solvent extraction. Informations on the basic studies carried out to ensure the reliability of the process are given in the complete paper

Phase transformation of metastable cubic γ-phase in U-Mo alloys

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Dey, G.K.; Kamath, H.S.

2010-01-01

Over the past decade considerable efforts have been put by many fuel designers to develop low enriched uranium (LEU 235 ) base U-Mo alloy as a potential fuel for core conversion of existing research and test reactors which are running on high enriched uranium (HEU > 85%U 235 ) fuel and also for the upcoming new reactors. U-Mo alloy with minimum 8 wt% molybdenum shows excellent metastability with cubic γ-phase in cast condition. However, it is important to characterize the decomposition behaviour of metastable cubic γ-uranium in its equilibrium products for in reactor fuel performance point of view. The present paper describes the phase transformation behaviour of cubic γ-uranium phase in U-Mo alloys with three different molybdenum compositions (i.e. 8 wt%, 9 wt% and 10 wt%). U-Mo alloys were prepared in an induction melting furnace and characterized by X-ray diffraction (XRD) method for phase determination. Microstructures were developed for samples in as cast condition. The alloys were hot rolled in cubic γ-phase to break the cast structure and then they were aged at 500 o C for 68 h and 240 h, so that metastable cubic γ-uranium will undergo eutectoid decomposition to form equilibrium phases of orthorhombic α-uranium and body centered tetragonal U 2 Mo intermetallic compound. U-Mo alloy samples with different ageing history were then characterized by XRD for phase and development of microstructure.

Microstructural evolution of a uranium-10 wt.% molybdenum alloy for nuclear reactor fuels

International Nuclear Information System (INIS)

Clarke, A.J.; Clarke, K.D.; McCabe, R.J.; Necker, C.T.; Papin, P.A.; Field, R.D.; Kelly, A.M.; Tucker, T.J.; Forsyth, R.T.; Dickerson, P.O.; Foley, J.C.; Swenson, H.; Aikin, R.M.; Dombrowski, D.E.

2015-01-01

Low-enriched uranium-10 wt.% molybdenum (LEU-10wt.%Mo) is of interest for the fabrication of monolithic fuels to replace highly-enriched uranium (HEU) dispersion fuels in high performance research and test reactors around the world. In this work, depleted uranium-10 wt.%Mo (DU-10wt.%Mo) is used to simulate the solidification and microstructural evolution of LEU-10wt.%Mo. Electron backscatter diffraction (EBSD) and complementary electron probe microanalysis (EPMA) reveal significant microsegregation present in the metastable γ-phase after solidification. Homogenization is performed at 800 and 1000 °C for times ranging from 1 to 32 h to explore the time–temperature combinations that will reduce the extent of microsegregation, as regions of higher and lower Mo content may influence local mechanical properties and provide preferred regions for γ-phase decomposition. We show for the first time that EBSD can be used to qualitatively assess microstructural evolution in DU-10wt.%Mo after homogenization treatments. Complementary EPMA is used to quantitatively confirm this finding. Homogenization at 1000 °C for 2–4 h may the regions that contain 8 wt.% Mo or lower, whereas homogenization at 1000 °C for longer than 8 h effectively saturates Mo chemical homogeneity, but results in substantial grain growth. The appropriate homogenization time will depend upon additional microstructural considerations, such as grain growth and intended subsequent processing. Higher carbon LEU-10wt.%Mo generally contains more inclusions within the grains and at grain boundaries after solidification. The effect of these inclusions on microstructural evolution (e.g. grain growth) during homogenization and as potential γ-phase decomposition nucleation sites is unclear, but likely requires additional study.

Microstructural evolution of a uranium-10 wt.% molybdenum alloy for nuclear reactor fuels

Energy Technology Data Exchange (ETDEWEB)

Clarke, A.J., E-mail: aclarke@lanl.gov; Clarke, K.D.; McCabe, R.J.; Necker, C.T.; Papin, P.A.; Field, R.D.; Kelly, A.M.; Tucker, T.J.; Forsyth, R.T.; Dickerson, P.O.; Foley, J.C.; Swenson, H.; Aikin, R.M.; Dombrowski, D.E.

2015-10-15

Low-enriched uranium-10 wt.% molybdenum (LEU-10wt.%Mo) is of interest for the fabrication of monolithic fuels to replace highly-enriched uranium (HEU) dispersion fuels in high performance research and test reactors around the world. In this work, depleted uranium-10 wt.%Mo (DU-10wt.%Mo) is used to simulate the solidification and microstructural evolution of LEU-10wt.%Mo. Electron backscatter diffraction (EBSD) and complementary electron probe microanalysis (EPMA) reveal significant microsegregation present in the metastable γ-phase after solidification. Homogenization is performed at 800 and 1000 °C for times ranging from 1 to 32 h to explore the time–temperature combinations that will reduce the extent of microsegregation, as regions of higher and lower Mo content may influence local mechanical properties and provide preferred regions for γ-phase decomposition. We show for the first time that EBSD can be used to qualitatively assess microstructural evolution in DU-10wt.%Mo after homogenization treatments. Complementary EPMA is used to quantitatively confirm this finding. Homogenization at 1000 °C for 2–4 h may the regions that contain 8 wt.% Mo or lower, whereas homogenization at 1000 °C for longer than 8 h effectively saturates Mo chemical homogeneity, but results in substantial grain growth. The appropriate homogenization time will depend upon additional microstructural considerations, such as grain growth and intended subsequent processing. Higher carbon LEU-10wt.%Mo generally contains more inclusions within the grains and at grain boundaries after solidification. The effect of these inclusions on microstructural evolution (e.g. grain growth) during homogenization and as potential γ-phase decomposition nucleation sites is unclear, but likely requires additional study.

Moessbauer study of defects in molybdenum and chromium irradiated with ions

International Nuclear Information System (INIS)

Troyan, V.A.; Bogdanov, V.V.; Ivanyushkin, E.M.; Pen'kov, Yu.P.

1980-01-01

Effects of ion irradiation of monocrystalline molybdenum and polycrystalline chromium with Co-57 impurity were studied by Moessbauer effect. Molybdenum specimens were irradiated by He + ions at accelerators with 40 keV energy. Chromium specimens were irradiated by hydrogen ions with 1.2 MeV energy up to integral 2x10 17 -2x10 19 ion/cm 2 doses. It is shown, that defect introduction into the source matrix by irradiation results in change of gamma-resonance line form and effect value. The observed effects of defect influence on spectrum parameters are discussed. It is concluded, that study of Moessbauer spectra parameters of diluted Co-57 solutions in matrices of different metals permits to determine dynamics of movement of impurity atoms and defects in metals irradiated with ions [ru

Study on segregation of aluminium-uranium alloys

International Nuclear Information System (INIS)

Lima, Rui Marques de

1979-01-01

The relations between alloy solidification and solute segregation were considered. The solidification structure and the solute redistribution during the solidification of alloys with dendritic micro morphology were studied. The macro and micro segregation theories were reviewed. The mechanisms that could change the solidification structure were taken into account in the context of more homogeneous alloy production. Aluminum alloys solidification structures and segregation were studied experimentally in the 13 to 45% uranium range, usually considering solidification in static molds. The uranium alloys with up to 20% uranium were studied both for solidification in ingot molds and for controlled directional solidification. It was verified that these alloy compositions had structures similar to those of hipoeutectic alloys, showing an a phase with dendritic morphology and inter dendritic eutectic. For the alloys with more than 25% uranium, it was observed the formation of UAl 3 and UAl 4 phases with dendritic morphology. The dendritic UAl 3 , phase morphology was affected both by the solute concentration in the alloy and by the growth rate. The dendritic UAl 3 phase non-singular aspect could be destroyed with decrease of the alloy solute concentration. In the alloys obtained with higher cooling rates it was found a tendency for the formation of substantial quantities of equi axial crystals of the solute enriched phases in the central regions of the ingot upper half. In the more external regions it was observed dendritic growth of these phases, for alloy compositions with over 25% uranium. An adequate reduction in the cooling rate changed the solidification structure form and distribution, as well as the segregation type and intensity. The uranium content in the solidified macro structures is presented as a function of: cooling rate, superheating, mold size, mold form and its temperature, number of remelting and time for the melt homogenization and agitation. It was

Corrosion resistant coatings for uranium and uranium alloys

International Nuclear Information System (INIS)

Weirick, L.J.; Lynch, C.T.

1977-01-01

Coatings to prevent the corrosion of uranium and uranium alloys are considered in two military applications: kinetic energy penetrators and aircraft counterweights. This study, which evaluated organic films and metallic coatings, demonstrated that the two most promising coatings are based on an electrodeposited nickel system and a galvanized zinc system

Development of silicide coating over molybdenum based refractory alloy and its characterization

International Nuclear Information System (INIS)

Chakraborty, S.P.; Banerjee, S.; Sharma, I.G.; Suri, A.K.

2010-01-01

Molybdenum based refractory alloys are potential candidate materials for structural applications in high temperature compact nuclear reactors and fusion reactors. However, these alloys being highly susceptible to oxidation in air or oxygen at elevated temperature, undergoes severe losses from highly volatile molybdenum trioxide species. Present investigation, therefore, examines the feasibility of development of silicide type of coating over molybdenum base TZM alloy shape (Mo > 99 wt.%) using pack cementation coating technique. TZM alloy was synthesized in this laboratory from oxide intermediates of MoO 2 , TiO 2 and ZrO 2 in presence of requisite amount of carbon, by alumino-thermic reduction smelting technique. The arc melted and homogenized samples of TZM alloy substrate was then embedded in the chosen and intimately mixed pack composition consisting of inert matrix (Al 2 O 3 ), coating powder (Si) and activator (NH 4 Cl) taken in the judicious proportion. The sealed charge packs contained in an alumina crucible were heated at temperatures of 1000 o C for 8-16 h heating cycle to develop the coating. The coating phase was confirmed to be of made of MoSi 2 by XRD analysis. The morphology of the coating was studied by SEM characterization. It had revealed that the coating was diffusion bonded where Si from coating diffused inward and Mo from TZM substrate diffused outward to form the coating. The coating was found to be resistant to oxidation when tested in air up to 1200 o C. A maximum 100 μm of coating thickness was achieved on each side of the substrate.

Irradiation experiment conceptual design parameters for MURR LEU U-Mo fuel conversion

International Nuclear Information System (INIS)

Stillman, J.; Feldman, E.; Stevens, J.; Wilson, E.

2013-03-01

This report contains the results of reactor design and performance calculations for conversion of the University of Missouri Research Reactor (MURR) 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) and the MURR Facility. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of its 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 nominal steady-state irradiation conditions of a key set of plates containing peak irradiation parameters found in MURR cores fueled with the LEU monolithic U-Mo alloy fuel with 10 wt% Mo.

Effect of molybdenum on structure, microstructure and mechanical properties of biomedical Ti-20Zr-Mo alloys.

Science.gov (United States)

Kuroda, Pedro Akira Bazaglia; Buzalaf, Marília Afonso Rabelo; Grandini, Carlos Roberto

2016-10-01

Titanium has an allotropic transformation around 883°C. Below this temperature, the crystalline structure is hexagonal close-packed (α phase), changing to body-centered cubic (β phase). Zirconium has the same allotropic transformation around 862°C. Molybdenum has body-centered cubic structure, being a strong β-stabilizer for the formation of titanium alloys. In this paper, the effect of substitutional molybdenum was analyzed on the structure, microstructure and selected mechanical properties of Ti-20Zr-Mo (wt%) alloys to be used in biomedical applications. The samples were prepared by arc-melting and characterized by x-ray diffraction with subsequent refinement by the Rietveld method, optical and scanning electron microscopy. The mechanical properties were analyzed by Vickers microhardness and dynamic elasticity modulus. X-ray measurements and Rietveld analysis revealed the presence of α' phase without molybdenum, α'+α″ phases with 2.5wt% of molybdenum, α″+β phases with 5 and 7.5wt% of molybdenum, and only β phase with 10wt% of molybdenum. These results were corroborated by microscopy results, with a microstructure composed of grains of β phase and lamellae and needles of α' and α″ phase in intra-grain the region. The hardness of the alloy was higher than the commercially pure titanium, due to the action of zirconium and molybdenum as hardening agents. The samples have a smaller elasticity modulus than the commercially pure titanium. Copyright © 2016 Elsevier B.V. All rights reserved.

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

Science.gov (United States)

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.

Irradiation of diffusion couples U-Mo/Al. Thermal calculation

International Nuclear Information System (INIS)

Fortis, Ana M.; Mirandou, Monica; Denis, Alicia C.

2004-01-01

The development of new low enrichment fuel elements for research reactors has lead to obtaining a number of compounds and alloys where the decrease in the enrichment is compensated by a higher uranium density in the fuel material. This has been achieved in particular with the uranium silicides dispersed in an aluminum matrix, where uranium densities about 4.8 g/cm 3 have been reached. Among the diverse candidate alloys, those of U-Mo with molybdenum content in the range 6 to 10 w % can yield, upon dispersion, to uranium densities of about 8 g/cm 3 . The first irradiation experiments employing these alloys in fuel plates, either dispersed in Al or monolithic revealed certain phenomena which are worthy of further studies. Failures have been detected apparently due to the formation of reaction products between the fissile material and the aluminum matrix, which exhibit a poor irradiation behavior. An experiment was designed which final purpose is to irradiate diffusion couples U-Mo/Al in the RA-3 reactor and to analyze the interaction zone at the working temperatures of the fuel elements. A simple device was built consisting of two Al 6063 blocks which press the U-Mo sample in between, located in an Al capsule. The ensemble is placed in a tube, which can be filled with different gases and introduced in the reactor. For safety reasons temperature predictions are necessary before performing the experiment. To this end, the COSMOS code was used. As a preliminary step and in order to test to exactness of the numerical estimations, two irradiations were performed in the RA-1 reactor with He and N 2 as transference gases. The agreement between the measured and calculated temperatures was good, particularly in the case of He and, along with the numerical predictions for the RA-3 reactor, provides a reliable basis to proceed with the following steps. (author)

Determination of U (Ⅵ) content in uranium molybdenum ores

International Nuclear Information System (INIS)

Wang Haisheng; Ding Hongfang

2014-01-01

Potentionmetric titration is established to determine U (Ⅵ) in uranium molybdenum ores. In the closed condition, U (Ⅵ) is leached by carbonate solution. U (Ⅵ) is reduced to U (Ⅳ) by ferrous sulfate in phosphoric acid. The exess ferrous sulfate is oxidized by sodium nitrite. urea decompose the exess sodium nitrite. U (Ⅳ) is titrated by ammonium metavanadate standard solution with potentionmetric titration. The precision is better than 5%, The recovery rate is 97.2%∼101.9%. (authors)

Transmission electron microscopy of oxide dispersion strengthened (ODS) molybdenum: effects of irradiation on material microstructure

International Nuclear Information System (INIS)

Baranwal, R.; Burke, M.G.

2003-01-01

Oxide dispersion strengthened (ODS) molybdenum has been characterized using transmission electron microscopy (TEM) to determine the effects of irradiation on material microstructure. This work describes the results-to-date from TEM characterization of unirradiated and irradiated ODS molybdenum. The general microstructure of the unirradiated material consists of fine molybdenum grains (< 5 (micro)m average grain size) with numerous low angle boundaries and isolated dislocation networks. 'Ribbon'-like lanthanum oxides are aligned along the working direction of the product form and are frequently associated with grain boundaries, serving to inhibit grain boundary and dislocation movement. In addition to the 'ribbons', discrete lanthanum oxide particles have also been detected. After irradiation, the material is characterized by the presence of nonuniformly distributed large (∼ 20 to 100 nm in diameter), multi-faceted voids, while the molybdenum grain size and oxide morphology appear to be unaffected by irradiation

Alloys under irradiation

International Nuclear Information System (INIS)

Martin, G.; Bellon, P.; Soisson, F.

1997-01-01

During the last two decades, some effort has been devoted to establishing a phenomenology for alloys under irradiation. Theoretically, the effects of the defect supersaturation, sustained defect fluxes and ballistic mixing on solid solubility under irradiation can now be formulated in a unified manner, at least for the most simple cases: coherent phase transformations and nearest-neighbor ballistic jumps. Even under such restrictive conditions, several intriguing features documented experimentally can be rationalized, sometimes in a quantitative manner and simple qualitative rules for alloy stability as a function of irradiation conditions can be formulated. A quasi-thermodynamic formalism can be proposed for alloys under irradiation. However, this point of view has limits illustrated by recent computer simulations. (orig.)

Gamma stability and powder formation of UMo alloys

Energy Technology Data Exchange (ETDEWEB)

Oliveira, F.B.V.; Andrade, D.A.; Angelo, G.; Belchior Junior, A.; Torres, W.M.; Umbehaun, P.E., E-mail: wmtorres@ipen.br, E-mail: umbehaun@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Angelo, E., E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, Sao Paulo, SP (Brazil). Grupo de Simulacao Numerica (GSN)

2015-07-01

A study of the hydrogen embrittlement as well as a research on the relation between gamma decomposition and powder formation of uranium molybdenum alloys were previously presented. In this study a comparison regarding the hypo-eutectoid and hyper-eutectoid molybdenum additions is presented. Gamma uranium molybdenum alloys have been considered as the fuel phase in plate type fuel elements for material and test reactors (MTR). Regarding their usage as a dispersion phase in aluminum matrix, it is necessary to convert the as cast structure into powder, and one of the techniques considered for this purpose is the hydration-dehydration (HDH). This paper shows that, under specific conditions of heating and cooling, γ-UMo fragmentation may occur with non-reactive or reactive mechanisms. Following the production of the alloys by induction melting, samples of the alloys were thermally treated under a constant flow of hydrogen. It was observed that, even without a massive hydration-dehydration process, the alloys fragmented under specific conditions of thermal treatment, during the thermal shock phase of the experiments. Also, there is a relation between absorption and the rate of gamma decomposition or the gamma phase stability of the alloy and this phenomenon can be related to the eutectoid transformation temperature. This study was carried out to search for a new method for the production of powders and for the evaluation of important physical parameter such as the eutectoid transformation temperature, as an alternative to the existing ones. (author)

Method of removing niobium from uranium-niobium alloy

International Nuclear Information System (INIS)

Pollock, E.N.; Schlier, D.S.; Shinopulos, G.

1992-01-01

This patent describes a method of removing niobium from a uranium-niobium alloy. It comprises dissolving the uranium-niobium alloy metal pieces in a first aqueous solution containing an acid selected from the group consisting of hydrochloric acid and sulfuric acid and fluoboric acid as a catalyst to provide a second aqueous solution, which includes uranium (U +4 ), acid radical ions, the acids insolubles including uranium oxides and niobium oxides; adding nitric acid to the insolubles to oxidize the niobium oxides to yield niobic acid and to complete the solubilization of any residual uranium; and separating the niobic acid from the nitric acid and solubilized uranium

Mechanical and irradiation properties of zirconium alloys irradiated in HANARO

International Nuclear Information System (INIS)

Kwon, Oh Hyun; Eom, Kyong Bo; Kim, Jae Ik; Suh, Jung Min; Jeon, Kyeong Lak

2011-01-01

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.1 10 21 n/cm 2 ). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed

Nanostructures obtained from a mechanically alloyed and heat treated molybdenum carbide

International Nuclear Information System (INIS)

Diaz Barriga Arceo, L.; Orozco, E.; Mendoza-Leon, H.; Palacios Gonzalez, E.; Leyte Guerrero, F.; Garibay Febles, V.

2007-01-01

Mechanical alloying was used to prepare molybdenum carbide. Microstructural characterization of samples was performed by X-ray diffraction (XRD), scanning electron microscopy (SEM) and transmission electron microscopy (TEM) methods. Molybdenum carbide was heated at 800 o C for 15 min in order to produce carbon nanotubes. Nanoparticles of about 50-140 nm in diameter and nanotubes with diameters of about 70-260 nm and 0.18-0.3 μm in length were obtained after heating at 800 o C, by means of this process

Nanostructures obtained from a mechanically alloyed and heat treated molybdenum carbide

Energy Technology Data Exchange (ETDEWEB)

Diaz Barriga Arceo, L. [Programa de Ingenieria Molecular, I.M.P. Lazaro Cardenas 152, C.P. 07730 D.F. Mexico (Mexico) and ESIQIE-UPALM, IPN Apdo Postal 118-395, C.P. 07051 D.F. Mexico (Mexico)]. E-mail: luchell@yahoo.com; Orozco, E. [Instituto de Fisica UNAM, Apdo Postal 20-364, C.P. 01000 D.F. Mexico (Mexico)]. E-mail: eorozco@fisica.unam.mx; Mendoza-Leon, H. [ESIQIE-UPALM, IPN Apdo Postal 118-395, C.P. 07051 D.F. Mexico (Mexico)]. E-mail: luchell@yahoo.com; Palacios Gonzalez, E. [Programa de Ingenieria Molecular, I.M.P. Lazaro Cardenas 152, C.P. 07730 D.F. Mexico (Mexico)]. E-mail: epalacio@imp.mx; Leyte Guerrero, F. [Programa de Ingenieria Molecular, I.M.P. Lazaro Cardenas 152, C.P. 07730 D.F. Mexico (Mexico)]. E-mail: fleyte@imp.mx; Garibay Febles, V. [Programa de Ingenieria Molecular, I.M.P. Lazaro Cardenas 152, C.P. 07730 D.F. Mexico (Mexico)]. E-mail: vgaribay@imp.mx

2007-05-31

Mechanical alloying was used to prepare molybdenum carbide. Microstructural characterization of samples was performed by X-ray diffraction (XRD), scanning electron microscopy (SEM) and transmission electron microscopy (TEM) methods. Molybdenum carbide was heated at 800 {sup o}C for 15 min in order to produce carbon nanotubes. Nanoparticles of about 50-140 nm in diameter and nanotubes with diameters of about 70-260 nm and 0.18-0.3 {mu}m in length were obtained after heating at 800 {sup o}C, by means of this process.

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

Science.gov (United States)

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

2002-12-01

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

Adhesive wear of iron chromium nickel silicon manganese molybdenum niobium alloys with duplex structure

International Nuclear Information System (INIS)

Lugscheider, E.; Deppe, E.; Ambroziak, A.; Melzer, A.

1991-01-01

Iron nickel chromium manganese silicon and iron chromium nickel manganese silicon molybdenum niobium alloys have a so-called duplex structure in a wide concentration range. This causes an excellent resistance to wear superior in the case of adhesive stress with optimized concentrations of manganese, silicon, molybdenum and niobium. The materials can be used for welded armouring structures wherever cobalt and boron-containing alloy systems are not permissible, e.g. in nuclear science. Within the framework of pre-investigations for manufacturing of filling wire electrodes, cast test pieces were set up with duplex structure, and their wear behavior was examined. (orig.) [de

TERNARY ALLOYS OF URANIUM, COLUMBIUM, AND ZIRCONIUM

Science.gov (United States)

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.

Contribution to the micrographic study of uranium and its alloys

International Nuclear Information System (INIS)

Monti, H.

1956-06-01

The present report is the result of research carried out by the radio metallurgy section, to perfect micrographic techniques applicable to the study of samples of irradiated uranium. In the first part of this work, two polishing baths are developed, having the qualities with a minimum of disadvantages inherent in their respective compositions: they are, on the one hand perchloric acid-ethanol mixtures, and on the other hand a phospho-chromic-ethanol bath. In the chapter following, the micrographic attack of uranium is studied. The only satisfactory process is oxidation by cathode bombardment forming epitaxic layers. In the third chapter, an attempt is made to characterise the different surface states of the uranium by dissolution potential measurements and electronic diffraction. In the fourth chapter are given some examples of the application of these techniques to the micrographic study of various uranium alloys. In an appendix, it is shown how the chemical oxidation after phospho-chromic-alcohol polishing allows the different inclusions present in the molten uranium to be distinguished. By X-ray diffraction, uranium monocarbide and mononitride inclusions in particular are characterised. (author) [fr

IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

Directory of Open Access Journals (Sweden)

M.K. MEYER

2014-04-01

Full Text Available 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. 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.

Irradiation performance of U-Mo monolithic fuel

Energy Technology Data Exchange (ETDEWEB)

Meyer, M. K.; Gan, J.; Jue, J. F.; Keiser, D. D.; Perez, E.; Robinson, A.; Wachs, D. M.; Woolstenhulme, N. [Idaho National Laboratory, Idaho (Korea, Republic of); Kim, Y.S.; Hofman, G. L. [Argonne National Laboratory, Lemont (United States)

2014-04-15

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. 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.

Chromatographic retention of molybdenum, titanium and uranium complexes for removal of some interferences in inductively-coupled plasma mass spectrometry

International Nuclear Information System (INIS)

Jiang, S.-J.; Palmieri, M.D.; Fritz, J.S.; Houk, R.S.; Iowa State Univ., of Science and Technology, Ames

1987-01-01

Complexes of molybdenum(VI) or titanium(IV) with N-methylfurohydroxamic acid (N-MFHA) are retained on a column packed with polystyrene/divinylbenzene. At the pH values chosen, copper, zinc and cadmium are washed rapidly through the column and are detected by inductively-coupled plasma mass spectrometry without interference from metal oxide ions of titanium or molybdenum. Detection limits are 1 to 2 μg l -1 , and analyte recoveries are essentially 100%. The resin capacity for the titanium and molybdenum complexes is sufficient for several hundred injections, and the complexes can be readily washed from the column. Uranium(VI) also forms a stable complex with N-MFHA, and ionization interference caused by excess of uranium can be avoided by chromatographic removal of the uranium complex. Various other potentially interfering elements with aqueous oxidation states of +4 or higher (e.g. Sn, W, Hf or Zr) could also be separated by this technique. 33 refs.; 4 figs.; 3 tabs

Matrix composition effects on the tensile properties of tungsten-molybdenum heavy alloys

International Nuclear Information System (INIS)

Bose, A.; German, R.N.

1990-01-01

Tungsten-base heavy alloys are liquid-phase sintered from mixed tungsten, nickel, and iron powders. The sintered product is a composite consisting of interlaced tungsten and solidified matrix (W-Ni-Fe) phases. These alloys are most useful in applications requiring high density, strength, and toughness. The design of improved tungsten heavy alloys has been the subject of several research investigations. Much success has taken place through improved processing, but parallel compositional studies have resulted in new microstructure-property combinations. As part of these investigations, the Ni/Fe ratio has been varied, with the general conclusion that optimal strength and ductility occur with a ratio between 2 and 4. Brittle intermetallic phases can form outside of this composition range. Historically, a 7/3 Ni/Fe ratio has been selected for processing studies. Recently, others reported higher ductilities and impact energies for 90 and 93 pct W heavy alloys with the 8/2 Ni/Fe ratio. Alternatively, these alloys can be strengthened by both solid solution and grain size refinement through incorporation of molybdenum, tantalum, or rhenium. These additions are soluble in both the tungsten and matrix phases and retard solution-reprecipitation during liquid phase sintering. In this study, the alloy composition was varied in the nickel/iron ratio and molybdenum was partially substituted for tungsten. The sintered tensile properties are assessed vs these compositional variations

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

Science.gov (United States)

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.

Investigation of the uranium-molybdenum diffusion in body centered {gamma} solid solutions; Etude de la diffusion uranium-molybdene dans la solution solide {gamma} cubique centree

Energy Technology Data Exchange (ETDEWEB)

Adda, Y; Mairy, C; Bouchet, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Philibert, J [IRSID, 78 - Saint-Germain-en-Laye (France)

1958-07-01

The body centered {gamma} phase uranium-molybdenum intermetallic diffusion has been studied by different technical methods: micrography, electronic microanalyser, microhardness. The values of several numbers of penetration coefficients are given, and their physical significations has been discussed. The diffusion coefficients, the frequency factor and activation energies has been determined for each concentration. After determination of the Kirkendall effect in this system, we calculated the intrinsic diffusion coefficient of uranium and molybdenum. (author) [French] La dilution intermetallique uranium-molybdene, en phase {gamma} cubique centree, a ete etudiee au moyen de differentes techniques: micrographie, microsonde electronique, microdurete. Les valeurs d'un certain nombre de coefficients de penetration sont donnees et leur signification physique discutee. Les coefficients de diffusion, les facteurs de frequence et les energies d'activation ont ete determines pour chaque concentration. Apres avoir mis en evidence un effet Kirkendall dans ce systeme, on a calcule les coefficients de diffusion intrinseques de l'uranium et du molybdene. (auteur)

Solidification microstructures of aluminium-uranium alloys

International Nuclear Information System (INIS)

Ambrozio Filho, F.; Vieira, R.R.

1976-01-01

The solidification of microstrutures of aluminium-uranium alloys in the range of 4 to 20% uranium is investigated. The solidification was obtained both in ingot molds and under controlled directional solidification. The conditions for the presence of primary crystals and eutectic are discussed and an analysis of the influence of variables (growth rate and thermal gradient in the liquid) on the alloy structure is made. The effect of cooling rate on the alloy structures has been determined. It is found that the resulting structure can be derived from the kinectics concept, as required by the coupled-zone theory. Suggestions on the qualitative intervals of composition and temperatures with eutectic growth are presented [pt

Texture in low-alloyed uranium alloys

International Nuclear Information System (INIS)

Sariel, J.

1982-08-01

The dependence of the preferred orientation of cast and heat-treated polycrystalline adjusted uranium and uranium -0.1 w/o chromium alloys on the production process was studied. The importance of obtaining material free of preferred orientation is explained, and a survey of the regular methods to determine preferred orientation is given. Dilatometry, tensile testing and x-ray diffraction were used to determine the extent of the directionality of these alloys. Data processing showed that these methods are insufficient in a case of a material without any plastic forming, because of unreproducibility of results. Two parameters are defined from the results of Schlz's method diffraction test. These parameters are shown theoretically and experimentally (by extreme-case samples) to give the deviation from isotropy. Application of these parameters to the examined samples showes that cast material has preferred orientation, though it is not systematic. This preferred orientation was reduced by adequate heat treatments

The development of uranium foil farication technology utilizing twin roll method for Mo-99 irradiation target

CERN Document Server

Kim, C K; Park, H D

2002-01-01

MDS Nordion in Canada, occupying about 75% of global supply of Mo-99 isotope, has provided the irradiation target of Mo-99 using the rod-type UAl sub x alloys with HEU(High Enrichment Uranium). ANL (Argonne National Laboratory) through co-operation with BATAN in Indonesia, leading RERTR (Reduced Enrichment for Research and Test Reactors) program substantially for nuclear non-proliferation, has designed and fabricated the annular cylinder of uranium targets, and successfully performed irradiation test, in order to develop the fabrication technology of fission Mo-99 using LEU(Low Enrichment Uranium). As the uranium foils could be fabricated in laboratory scale, not in commercialized scale by hot rolling method due to significant problems in foil quality, productivity and economic efficiency, attention has shifted to the development of new technology. Under these circumstances, the invention of uranium foil fabrication technology utilizing twin-roll casting method in KAERI is found to be able to fabricate LEU or...

Effect of temperature on the crack resistance of a molybdenum alloy with 30% tungsten

International Nuclear Information System (INIS)

Uskov, E.I.; Babak, A.V.; D'yachkov, A.P.; Platonov, V.A.

1986-01-01

Results are presented for a study of the effect of temperature on the crack resistance of molybdenum alloy with 30% tungsten (Mo - 30% W), and data are presented for the crack resistance of commercial-purity molybdenum and tungsten obtained by power metallurgy in the temperature range 20-1800 C. It was found that the nature of failure for Mo-30% W alloy depends on test temperature; in the temperature range 20 C-T /SUP d/ /SUB br/ (upper boundary for the temperature range of the ductile to brittle transition), failure is unstable in nature, and at temperatures exceeding this transition, it occurs by steady main crack development

Effect of molybdenum and chromium additions on the mechanical properties of Fe3Al-based alloys

International Nuclear Information System (INIS)

Sun Yangshan; Xue Feng; Mei Jianping; Yu Xingquan; Zhang Lining

1995-01-01

Iron aluminides based on Fe 3 Al offer excellent oxidation and sulfidation resistance, with lower material cost and density than stainless steels. However, their potential use as structural material has been hindered by limited ductility and a sharp drop in strength above 600 C. Recent development efforts have indicated that adequate engineering ductility of 10--20% and tensile yield strength of as high as 500 MPa can be achieved through control of composition and microstructure. These improved tensile properties make Fe 3 Al-based alloys more competitive against conventional austenic and ferritic steels. The improvement of high temperature mechanical properties has been achieved mainly by alloying processes. Molybdenum has been found to be one of the most important alloying elements for strengthening Fe 3 Al-based alloys at high temperatures. However, the RT(room temperature) ductility decreases with the increase of a molybdenum addition. On the other hand, a chromium addition to Fe 3 Al-based alloys is very efficient for improving RT ductility but not beneficial to yield strength at temperatures to 800 C. The purpose of the present paper is to report the effects of combined additions of molybdenum and chromium on mechanical properties at ambient temperature and high temperature of 600 C

Localized corrosion of molybdenum-bearing nickel alloys in chloride solutions

International Nuclear Information System (INIS)

Postlethwaite, J.; Scoular, R.J.; Dobbin, M.H.

1988-01-01

Electrochemical and immersion tests have been applied to a study of the localized corrosion resistance of two molybdenum-bearing nickel alloys. Alloys C-276 and 6y25, in neutral chloride solutions in the temperature range of 25 to 200 C as part of the container materials evaluation screening tests for the Canadian Nuclear Fuel Waste Management Program. Cyclic polarization studies show that the passivation breakdown potentials move rapidly to more active values with increasing temperatures, indicating a reduced resistance to localized corrosion. The results of immersion tests show that both alloys do suffer crevice corrosion in neutral aerated sodium chloride solutions at elevated temperatures, but that in both cases there is a limiting temperature > 100C, below which, the alloys are not attacked, regardless of the chloride concentration

Study of the recrystallisation of irradiated uranium

International Nuclear Information System (INIS)

Bloch, J.; Mustelier, J.P.; Bussy, P.; Blin, J.

1958-01-01

1- Study of the recrystallisation of irradiated uranium. The recrystallisation of uranium irradiated to a burnup level of 220 MWj/t, at a temperature of the order of 350 deg. C, has been investigated. The observations were made chiefly by means of micrography an hardness measurements. If the irradiated metal is compared with a cold-drawn metal showing the same shearing of the twinned crystals, and therefore the same rate of plastic deformation, as the irradiated metal, it is noted that the restoring of the irradiated metal takes place at a considerably higher temperature than that of the cold-drawn metal. Pre-crystallisation is very much delayed. Only, a passage of the α-β transformation point quickly wipes out irradiation effect. 2- Hardening of uranium by irradiation. Using hardness measurements we have studied more especially the effect of very weak irradiations on uranium (integrated flux 16 nvt). The hardness does not increase linearly with the flux, but a period of incubation is observed probably representing the time necessary for saturation of the dislocations. (author) [fr

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

Energy Technology Data Exchange (ETDEWEB)

Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

2011-02-01

This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

Orientational relationships between phases in the γ→α transformations for uranium-molybdenum alloys

International Nuclear Information System (INIS)

Brun, G.

1966-04-01

A crystallographic study has been made of the γ → α + γ transformation in the alloy containing 3 per cent by weight of molybdenum using electronic micro-diffraction; it has been possible to establish the orientational relationships governing the germination of the α phase in the γ phase. One finds: (111)γ // (100) α, (112-bar)γ // (010) α, (11-bar 0)γ // (001)α. By choosing a monoclinic lattice containing the same number of atoms as the orthorhombic lattice for defining the γ mother phase, the change in structure has been explained by adding a homogeneous (112-bar)γ [111]γ shearing deformation to a heterogeneous deformation brought about by slipping of the atoms which are not situated at the nodes of this lattice. The identity of the orientation relationships γ/α and γ/α''b and the loss of coherence γ /α as a function of temperature or of time lead to the conclusion that, in the range studied, the γ → α transformation begins with a martensitic process and continues by germination and growth. (author) [fr

Alloys of uranium and aluminium with low aluminium content

International Nuclear Information System (INIS)

Cabane, G.; Englander, M.; Lehmann, J.

1955-01-01

Uranium, as obtained after spinning in phase γ, presents an heterogeneous structure with large size grains. The anisotropic structure of the metal leads to an important buckling and surface distortion of the fuel slug which is incompatible with its tubular cladding for nuclear fuel uses. Different treatments have been made to obtain an isotropic structure presenting high thermal stability (laminating, hammering and spinning in phase α) without success. Alloys of uranium and aluminium with low aluminium content present important advantage in respect of non allied uranium. The introduction of aluminium in the form of intermetallic compound (UAl 2 ) gives a better resistance to thermal fatigue. Alloys obtained from raw casting present an improved buckling and surface distortion in respect of pure uranium. This improvement is obtained with uranium containing between 0,15 and 0,5 % of aluminium. An even more improvement in thermal stability is obtained by thermal treatments of these alloys. These new characteristics are explained by the fine dispersion of the UAl 2 particles in uranium. The results after treatments obtained from an alloy slug containing 0,4 % of aluminium show no buckling or surface distortion and no elongation. (M.P.)

Molybdenum isotope fractionation during acid leaching of a granitic uranium ore

Science.gov (United States)

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

Shape memory effects in a uranium + 14 at. % niobium alloy

International Nuclear Information System (INIS)

Vandermeer, R.A.; Ogle, J.C.; Snyder, W.B. Jr.

1978-01-01

There is a class of alloys that, on cooling from elevated temperatures, experience a martensitic phase change. Some of these, when stressed in the martensitic state to an apparently plastic strain, recover their predeformed shape simply by heating. This striking shape recovery is known as the ''shape memory effect'' (SME). Up to a certain limiting strain, epsilon/sub L/, 100% shape recovery may be accomplished. This memory phenomenon seems to be attributable to the thermoelastic nature of and deformational modes associated with the phase transformation in the alloy. Thus, shape recovery results when a stress-biased martensite undergoes a heat-activated reversion back to the parent phase from which it originated. There are uranium alloys that demonstrate SME-behavior. Uranium-rich, uranium--niobium alloys were the first to be documented; New experimental observations of SME in a polycrystalline uranium--niobium alloy are presented. This alloy can exhibit a two-way memory under cetain circumstances. Additional indirect evidence is presented suggesting that the characteristics of the accompanying phase transformation in this alloy meet the criteria or ''selection rules'' deemed essential for SME

Major constituent quantitative determination in uranium alloys by coupled plasma atomic emission spectrometry and X ray fluorescence wavelength dispersive spectrometry

International Nuclear Information System (INIS)

Oliveira, Luis Claudio de; Silva, Adriana Mascarenhas Martins da; Gomide, Ricardo Goncalves; Silva, Ieda de Souza

2013-01-01

A wavelength-dispersive X-ray fluorescence (WD-XRF) spectrometric method for determination of major constituents elements (Zr, Nb, Mo) in Uranium/Zirconium/Niobium and Uranium/Molybdenum alloy samples were developed. The methods use samples taken in the form of chips that were dissolved in hot nitric acid and precipitate particles melted with lithium tetraborate and dissolved in hot nitric acid and finally analyzed as a solution. Studies on the determination by inductively coupled plasma optic emission spectrometry (ICP OES) using matched matrix in calibration curve were developed. The same samples solution were analyzed in both methods. The limits of detection (LOD), linearity of the calibrations curves, recovery study, accuracy and precision of the both techniques were carried out. The results were compared. (author)

Fluorimetric determination of uranium in zirconium and zircaloy alloys

International Nuclear Information System (INIS)

Acosta L, E.

1991-05-01

The objective of this procedure is to determine microquantities of uranium in zirconium and zircaloy alloys. The report also covers the determination of uranium in zirconium alloys and zircaloy in the range from 0.25 to 20 ppm on 1 g of base sample of radioactive material. These limit its can be variable if the size of the used aliquot one is changed for the final determination of uranium. (Author)

Passivation and corrosion behaviours of cobalt and cobalt-chromium-molybdenum alloy

International Nuclear Information System (INIS)

Metikos-Hukovic, M.; Babic, R.

2007-01-01

Passivation and corrosion behaviour of the cobalt and cobalt-base alloy Co30Cr6Mo was studied in a simulated physiological solution containing chloride and bicarbonate ions and with pH of 6.8. The oxido-reduction processes included solid state transformations occurring at the cobalt/electrolyte interface are interpreted using theories of surface electrochemistry. The dissolution of cobalt is significantly suppressed by alloying it with chromium and molybdenum, since the alloy exhibited 'chromium like' passivity. The structural and protective properties of passive oxide films formed spontaneously at the open circuit potential or during the anodic polarization were studied using electrochemical impedance spectroscopy in the wide frequency range

Change in mechanical properties of low-alloyed molybdenum alloys at two-stage strengthening during aging

International Nuclear Information System (INIS)

Bernshtejn, L.M.; Zakharov, A.M.; Arbuzov, V.K.

1977-01-01

Change in mechanical properties of hardened low-alloyed molybdenum alloys (Mo-Zr-C and Mo-Zr-Nb-C) at two-stage strengthening during ageing at 1400 deg C is studied. The initial strengthening maximum following ageing for 5 hr is caused by separation of dispersed ZrC particles and is accompanied by worsened plasticity, a development characteristic of precipitation hardening processes. The second increase in strength after a 10-hr ageing is not accompanied by reduced plasticity, this being characteristic of strengthening as a result of reconstruction of the dislocation structure. Niobium (0.16 wt.%) added to Mo-Zr-C alloys simultaneously increases their plastic and strength properties. The said effect is caused by prevention of premature decomposition of alloys on account of increased low-temperature plasticity, which permits to obtain high resistance to plastic deformation

Wetting of molybdenum with molten Cu-O alloys

International Nuclear Information System (INIS)

Yupko, V.L.; Garbuz, V.V.; Kryuchkova, N.I.

1992-01-01

The Cu-O alloys were prepared from type MOb copper (GOST 859-78) with an oxygen content of 0.001 wt.% and type ChDA cuprous oxide (MRTU 6-09-1451-64), the powder of which was first pressed into briquettes. The weighted portions of Cu 2 O were weighed on an Elektrobalans scale having an absolute error of ±5 · 10 -7 g. The relative error in weighing an approximately 1 · 10 -4 g weighed portion of Cu 2 O for preparation of the alloy with the minimum oxygen content of 0.002% was, therefore, ± 0.5% and consequently for the alloys with a higher oxygen content the accuracy was higher. The alloys were prepared on a ZrO 2 + 5% Y 2 O 3 ceramic at 1,420 K in a vacuum of 6.7 · 10 -3 Pa,d their weight was 1.0-1.5 g, and the melting time 30 sec. The pure type MOb copper was remelted in the same manner. The time relationships of the angle of wetting of molybdenum by molten Cu-O alloys under conditions of combined heating are given. With an increase in oxygen content from 0.004 to 0.005%, wetting drops sharply

Procedure for Uranium-Molybdenum Density Measurements and Porosity Determination

Energy Technology Data Exchange (ETDEWEB)

Prabhakaran, Ramprashad [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Devaraj, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

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.

X-ray topography of uranium alloys

International Nuclear Information System (INIS)

Le Naour, L.

1984-01-01

The limitations of x-ray topography methods are due to the variety of structures studied and to the variation of the amplitude of the scattering of incident beams. It is difficult to evaluate the aberrations and the imperfections of the material studied. Interpretation of the x-ray images will often be delicate and that is aggravated by the complexity of the diffraction spectrum of uranium. This negative aspect is compensated for by the advantage that chemical or electrochemical preparations of the alloy surface, along with alterations that can take place and the lack of trueness are avoided. Precise and very reproducible numerical data can be derived from the patterns. The structure of alloys, at a given scale, is revealed and characterized by quantitative parameters such as size of grains or sub-grains, dispersion of their dimensions, mutual disorientations and the continuous or discontinuous nature of the latter. The results of this research, therefore, justify the use of methods inspired by the Berg-Barrett technique. These diffraction procedures constitute a useful means for investigating many elements of microstructure that closely govern the behavior under irradiation of the materials being examined

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

Energy Technology Data Exchange (ETDEWEB)

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

2016-10-01

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, and 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.

Determination of hafnium, molybdenum, and vanadium in niobium and niobium-based alloys by atomic absorption spectrometry

International Nuclear Information System (INIS)

Ide, Kunikazu; Kobayashi, Takeshi; Sudo, Emiko.

1985-01-01

The analytical procedure is as follows: Weigh 1 g of a sample and put it into a 100 cm 3 PTFE beaker. Add 5 ml of distilled water and 5 ml of hydrofluoric acid, and then heat the solution on a hot plate, adding 3 ml of nitric acid dropwise. Dilute the solution to 100 cm 3 with distilled water. When hafnium is determined, add 2 g of diammonium titanium hexafluoride ((NH 4 ) 2 TiF 6 )) before dilution. Working standard solutions are prepared by adding the stock standard solutions of hafnium, molybdenum, and vanadium into niobium solutions. When hafnium is determined, add 2 g of (NH 4 ) 2 TiF 6 and the alloying elements in amounts corresponding to those in sample solutions into the working standard solutions. The tolerable amounts of hydrofluoric acid were 2.9 M, 2.1 M, and 3.1 M and those of nitric acid were 1.0 M, 1.6 M, and 1.6 M for hafnium, molybdenum, and vanadium, respectively. It was found that (NH 4 ) 2 TiF 6 greatly increased the sensitivity for hafnium determination. Niobium showed minus effect for hafnium and plus effect for molybdenum and vanadium. The atomic absorption of molybdenum and vanadium were not influenced by the presence of 20 % of each alloying element, while the atomic absorption of hafnium was given plus effect by 20 % of zirconium, iron, cobalt, nickel, manganese, chromium or vanadium and minus effect by 20 % tungsten. The analytical values of hafnium, molybdenum, and vanadium in niobium-based alloys by this method showed a good agreement with those by X-ray fluorescence analysis. The lower limits of determination (S/N=2) were 0.05, 0.001, and 0.002 % and the relative standard deviation were 3, 1, and 1.5 % for hafnium, molybdenum, and vanadium, respectively. (author)

Polarographic methods for the analysis of beryllium metal and its alloys

International Nuclear Information System (INIS)

Wells, J.M.

1975-10-01

This report describes polarographic methods for the analysis of beryllium metal and its alloys. The elements covered by these methods are aluminium, bismuth, cadmium, cobalt, copper, iron, lead, molybdenum, nickel, thallium, tungsten, uranium, vanadium and zinc. (author)

Study of the oxidation kinetics of the nickel-molybdenum alloy

International Nuclear Information System (INIS)

Gouillon, Marie-Josephe

1974-01-01

This research thesis reports the study of the oxidation of a nickel-molybdenum alloy in the high-nickel-content part of this alloy. After a bibliographical study on the both metals, the author proposes a physical model based on observed phenomena and based on experimental results. Based on a thermodynamic study, the author compares the stability of the different oxides which may be formed, and reports a prediction of oxides obtained on the alloy during oxidation. Qualitative and quantitative studies have been performed by scanning electron microscopy coupled with electronic microprobe analysis to investigate morphological characteristics on oxidation films. A kinetic study by thermogravimetry shows a decrease of the alloy oxidation rate with respect to that of pure nickel at temperatures lower than 800 degrees C. This result is interpreted by the intervention of two opposed diffusion phenomena which act against each other [fr

Evaluation of AS-CAST U-Mo alloys processed in graphite crucible coated with boron nitride

Energy Technology Data Exchange (ETDEWEB)

Marra, Kleiner M., E-mail: kleiner.marra@prof.una.br [Centro Universitario UNA, Belo Horizonte, MG (Brazil). Curso de Engenharia Mecânica; Reis, Sérgio C.; Paula, João B. de; Pedrosa, Tércio A., E-mail: reissc@cdtn.br, E-mail: jbp@cdtn.br, E-mail: tap@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

This paper reports the production of uranium-molybdenum alloys, which have been considered promising fuel for test and research nuclear reactors. U-Mo alloys were produced in three molybdenum contents: 5w%, 7w%, and 10w%, using an electric vacuum induction furnace. A boron nitride-coated graphite crucible was employed in the production of the alloys and, after melting, the material was immediately poured into a boron nitride-coated graphite mold. The incorporation of carbon was observed, but it happened in a lower intensity than in the case of the non-coated crucible/mold. It is observed that the carbon incorporation increased and alloys density decreased with Mo addition. It was also noticed that the increase in the carbon or molybdenum content did not seem to change the as-cast structure in terms of granulation. The three alloys presented body-centered cubic crystal structure (γ-phase), after solidification, besides a seeming negative microsegregation of molybdenum, from the center to the periphery of the grains. There were signs of macrosegregation, from the base to the top of the ingots. (author)

Irradiated uranium reprocessing; Prerada ozracenog urana

Energy Technology Data Exchange (ETDEWEB)

Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorijaza visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

Task concerned with reprocessing of irradiated uranium covered the following activities: implementing the method and constructing the cell for uranium dissolving; implementing the procedure for extraction of uranium, plutonium and fission products from radioactive uranium solutions; studying the possibilities for using inorganic ion exchangers and adsorbers for separation of U, Pu and fission products.

Copper, zinc, molybdenum and uranium distribution in bottom sediments of the Black Sea

International Nuclear Information System (INIS)

Zhorov, V.A.; Sovga, E.E.; Solov'eva, L.B.; Oguslavskij, P.G.; Babinets, A.E.; AN Ukrainskoj SSR, Kiev. Inst. Geologicheskikh Nauk)

1983-01-01

The results of investigations of bottom sediments of the Black Sea by four expeditions aboard scientific ships ''Academician Vernadsky'', ''Michael Lomonosov'', ''Academician Vavilov'' in 1972-1978, are presented. 70 columns of bottom sediments are studied, about 200 samples are analyzed for Cu, Zn, Mo and U using chemical methods with photometric ending. Charts of Cu, Zn, Mo and U distribution in modern, ancient Black Sea and neoeuxenic sediments of the basin are prepared. Preferable uranium concentration in modern sediments, copper and molybdenum - in sapropelic muds of ancient Black Sea sediments and zinc - in neoeuxenic layers, is shown. Uranium geochemical behaviour is determined by physico-chemical regime of the basin, the presence of restoring situation which promotes the formation of uranium sorption-active forms in the upper layer of modern sediments. Neither sapropelite (organic matter), nor the peculiarities of lithological composition of sediments affect uranium behaviour

Features of soldering of molybdenum a lols

International Nuclear Information System (INIS)

Grishin, V.L.; Rybkin, B.V.; Cherkasov, A.F.

1980-01-01

Soldering features of complex-alloy molybdenum alloys were investigated in comparison with alloys based on solid solutions. Soldering features of heterogeneous molybdenum base alloys were investigated using samples of 0.5-1.O mm sheets with the strain of about 95% made of ingots which had been smelted in arc vacuum furnaces. The soldering of samples was carried out in 5x1O -5 mm Hg vacuum using different sources of heating: radiation, electron-ray and contact. It was shown that heat-resisting soldered joints of heterogeneous molybdenum alloys could be produced using zirconium and niobium base solders containing the most effective hardeners of the parent material (titanum, vanadium, tantalum, molybdenum, tungsten). To preserve high mechanical properties of heterogeneous alloys it was expedient to use for welding local heating sources which permitted to decrease considerably temperature- time conditions of the process

Damage production by fast electrons in dilute alloys of vanadium, niobium and molybdenum

International Nuclear Information System (INIS)

Jung, P.

1975-01-01

Vanadium, niobium and molybdenum samples containing about 300 ppm of zirconium were irradiated at helium temperature with electrons of energies between 0,6 and 3.1 MeV. The measured damage rates were analysed in terms of minimum threshold energy, damage function and resistivity per unit concentration of Frenkel pairs. For the minimum threshold energy T(Sub)d, values of 25+-2 eV (V) 28+-2 e V(Nb) and 34+-2 e V(Mo) were obtained. Pronounced differences between the displacement functions of molybdenum and that of niobium and vanadium are found which are explained by different stability of the defects during the irradiation at helium temperature

Welding of a powder metallurgy uranium alloy

International Nuclear Information System (INIS)

Holbert, R.K.; Doughty, M.W.; Alexander-Morrison, G.M.

1989-01-01

The interest at the Oak Ridge Y-12 Plant in powder metallurgy (P/M) uranium parts is due to the potential cost savings in the fabrication of the material, to achieving a more homogeneous product, and to the reduction of uranium scrap. The joining of P/M uranium-6 wt-% niobium (U-6Nb) alloys by the electron beam (EB) welding process results in weld porosity. Varying the EB welding parameters did not eliminate the porosity. Reducing the oxygen and nitrogen content in this P/M uranium material did minimize the weld porosity, but this step made the techniques of producing the material more difficult. Therefore, joining wrought and P/M U-6Nb rods with the inertia welding technique is considered. Since no gases will be evolved with the solid-state welding process and the weld area will be compacted, porosity should not be a problem in the inertia welding of uranium alloys. The welds that are evaluated are wrought-to-wrought, wrought-to-P/M, and P/M-to-P/M U-6Nb samples

Self-irradiation study of plutonium alloys

International Nuclear Information System (INIS)

Oudot, B.

2005-02-01

The plutonium is unstable and produces α or β decays depending on the isotope. These decays generate americium, uranium, helium and different kinds of structural defects. The effects of self-irradiation damage are observed at macroscopic scale, the mechanism occurs from atomic scale. In order to improve our understanding of the self-irradiation effects in PuGa alloys, a technique sensitive to the vacancies and vacancies clusters has been developed: the Positron Annihilation Spectroscopy (PAS). The swelling has been characterized by XRD at a microscopic scale and by dilatometry at a macroscopic scale. Swelling starts just after melting and reaches a saturation between 6 and 36 months depending on the degree of gallium homogeneity in the alloy. Swelling at saturation increases with the gallium content, but the absolute change in the cell parameters is constant during time. PAS showed that vacancies clusters develop immediately. Their concentration increase with time. A part of these clusters is stabilized by helium atoms and leads to the creation of bubbles, which contribution to swelling is negligible. The vacancies and vacancies clusters which are not stabilized by helium contribute to the swelling increase by mechanisms known for other materials. These mechanisms are based on a 'dislocation bias'. The presence of these dislocations can furthermore explain the low mean life time value of positrons at the saturation point. (author)

The reprocessing of irradiated fuels improvement and extension of the solvent extraction process; Le traitement des combustibles irradies amelioration et extension du procede utilisant les solvants

Energy Technology Data Exchange (ETDEWEB)

Faugeras, P; Chesne, A [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

Improvements made in the conventional tri-butylphosphate process are described, in particular. the concentration and the purification of plutonium by one extraction cycle using tri-butyl-phosphate with reflux; and the use of an apparatus working continuously for precipitating plutonium oxalate, for calcining the oxalate, and for fluorinating the oxide. The modifications proposed for the treatment of irradiated uranium - molybdenum alloys are described, in particular, the dissolution of the fuel, and the concentration of the fission product solutions. The solvent extraction treatment is used also for the plutonium fuels utilized for the fast breeder reactor (Rapsodie) An outline of the process is presented and discussed, as well as the first experimental results and the plans for a pilot plant having a capacity of 1 kg/day. The possible use of tn-lauryl-amine in the plutonium purification cycle is now under consideration for the processing plant at La Hague. The flowsheet for this process and its performance are presented. The possibility of vitrification is considered for the final treatment of the concentrated radioactive wastes from the Marcoule (irradiated uranium) and La Hague (irradiated uranium-molybdenum) Centers. Three possible processes are described and discussed, as well as the results obtained from the operation of the corresponding experimental units using tracers. (authors) [French] On decrit les ameliorations apportees au procede classique utilisant le phosphate tributylique, et notamment la concentration et la purification du plutonium par un cycle d'extraction au tributylphosphate avec reflux, l'utilisation d'un appareillage continu de precipitation d'oxalate de plutonium, de calcination de l'oxalate, et de fluoration de l'oxyde. On presente les modifications envisagees pour le traitement des alliages uranium-molybdene irradies, principalement en ce qui concerne la dissolution du combustible et la concentration des solutions de produits de fission

Production of fusion radionuclides: Molybdenum-99/ Iodine - 131 and Xenon-133

International Nuclear Information System (INIS)

Barrachina, M.; Carrillo, D.

1982-01-01

This report presents a new radiochemical method for industrial production of the radionuclides: molybdenum-99, iodine-131 and xenon-133. The above mentioned method based on the alkaline metathesis reaction of irradiated uranium (IV) fluoride, presents the best characteristics for the proposed objective. The study deals with the analysis of that reaction and the separation and purification processes. (Author) 71 refs

Durability of adhesive bonds to uranium alloys, tungsten, tantalum, and thorium

International Nuclear Information System (INIS)

Childress, F.G.

1975-01-01

Long-term durability of epoxy bonds to alloys of uranium (U-Nb and Mulberry), nickel-plated uranium, thorium, tungsten, tantalum, tantalum--10 percent tungsten, and aluminum was evaluated. Significant strengths remain after ten years of aging; however, there is some evidence of bond deterioration with uranium alloys and thorium stored in ambient laboratory air

High strength ferritic alloy

International Nuclear Information System (INIS)

1977-01-01

A high strength ferritic steel is specified in which the major alloying elements are chromium and molybdenum, with smaller quantities of niobium, vanadium, silicon, manganese and carbon. The maximum swelling is specified for various irradiation conditions. Rupture strength is also specified. (U.K.)

Alloys of uranium and aluminium with low aluminium content; Alliages uranium-aluminium a faible teneur en aluminium

Energy Technology Data Exchange (ETDEWEB)

Cabane, G; Englander, M; Lehmann, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

Uranium, as obtained after spinning in phase {gamma}, presents an heterogeneous structure with large size grains. The anisotropic structure of the metal leads to an important buckling and surface distortion of the fuel slug which is incompatible with its tubular cladding for nuclear fuel uses. Different treatments have been made to obtain an isotropic structure presenting high thermal stability (laminating, hammering and spinning in phase {alpha}) without success. Alloys of uranium and aluminium with low aluminium content present important advantage in respect of non allied uranium. The introduction of aluminium in the form of intermetallic compound (UAl{sub 2}) gives a better resistance to thermal fatigue. Alloys obtained from raw casting present an improved buckling and surface distortion in respect of pure uranium. This improvement is obtained with uranium containing between 0,15 and 0,5 % of aluminium. An even more improvement in thermal stability is obtained by thermal treatments of these alloys. These new characteristics are explained by the fine dispersion of the UAl{sub 2} particles in uranium. The results after treatments obtained from an alloy slug containing 0,4 % of aluminium show no buckling or surface distortion and no elongation. (M.P.)

Spectrographic analysis of uranium-based alloys; Analyse spectrographique d'alliages a base d'uranium

Energy Technology Data Exchange (ETDEWEB)

Baudin, G.; Blum, P.

1959-07-01

The authors describe a spectrographic method for dosing cobalt in cobalt-uranium alloys with cobalt content from 0.05 to 10 per cent. They describe sample preparation, alloy solution, spectrographic conditions, and photometry operations. In a second part, they address the dosing of boron in uranium borides. They implement the so-called 'porous cup' method. Boride is dissolved by fusion with Co{sub 3}-NaK [French] Uranium-Cobalt: il est decrit une methode spectrographique de dosage de cobalt dans des alliages cobalt-uranium pour des teneurs de 0,05 pour cent a 10 pour cent de Co. On opere sur solution avec le fer comme standard interne. Borure d'Uranium: ici encore on opere par la methode dite 'porous cup', le fer etant conserve comme standard interne. Le borure est mis en solution par fusion avec Co{sub 3}NaK. (auteurs)

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

International Nuclear Information System (INIS)

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 27 Al(p,γ) 28 Si reaction which occurs at a proton energy of 991.9 keV. During irradiation at 150 deg. C, aluminum diffused into the uranium alloy at a irradiation driven diffusion rate of 30 nm 2 /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

Relationship between uranium-molybdenum, fluorite and gold deposits within provinces of continental volcanicity

International Nuclear Information System (INIS)

Modnikov, I.S.; Skvortsova, K.V.; Chesnokov, L.V.

1974-01-01

The article gives a comparative description of and the age relationships between uranium-molybdenum, gold and fluorite mineralizations in the areas of development of adhesite-diorite and liparite-granite vulcanoplutonic formations, which are most fully and intensively manifest in the intra-anticlinal and median blocks of folded regions in the final stages of geosynclinal development or during the final stages of tectono-magmatic activation. These formations usually fill vulcano-tectonic depression structures - overlaid troughs and inherited delections. The geological and geochemical data are evidence of the close temporal link between the hydrothermal process of ore formation and the type and scale of manifestations of the vulcano-plutonic magmatism that is responsible for the general geochemical features of the ores of deposits of various types. The formation of gold, fluorite and uranium-molybdenum deposits occurred immediately after the completion of effusive and intrusive magmatism during a single metallogenic cycle. The spatial distribution of the ore fields and deposits depends chiefly on the peculiarities of the tectonic make-up of the depression structures, and also on the type and scale of the manifestations of vulcano-plutonic magmatism. (B.Ya.)

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

Science.gov (United States)

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

Study of the pyrophoric characteristics of uranium-iron alloys

International Nuclear Information System (INIS)

Duplessis, X.

2000-01-01

The objective of the study is to understand the pyrophoric characteristics of uranium-iron alloys. In order to carry out this research we have elected to use uranium-iron alloy powder with granules of 200 μm and 1000 μm diameter with 4%, 10.8% and 14% iron content. The experiments were performed on small samples of few milligrams and on larger quantities of few hundred grams. The main conclusions obtained are the followings: -The reaction start at 453 K (180 deg. C) and the ignition at 543 K (270 deg. C) - The influence of the specific area seems more important than the iron concentration in the alloys - When the alloy ignites, the fire spreads quickly and the alloy rapidly consumes. (author)

Experimental study on uranium alloys for hydrogen storage

International Nuclear Information System (INIS)

Deaconu, M.; Meleg, T.; Dinu, A.; Mihalache, M.; Ciuca, I.; Abrudeanu, M.

2013-01-01

The heaviest isotope of hydrogen is one of critically important elements in the field of fusion reactor technology. Conventionally, uranium metal is used for the storage of heavier isotopes of hydrogen (D and T). Under appropriate conditions, uranium absorbs hydrogen to form a stable UH 3 compound when exposed to molecular hydrogen at the temperature range of 300-500 O C at varied operating pressure below one atmosphere. However, hydriding-dehydriding on pure uranium disintegrates the specimen into fine powder. The powder is highly pyrophoric and has low heat conductivity, which makes it difficult to control the temperature, and has a high possibility of contamination Due to the powdering effect as hydrogen in uranium, alloying uranium with other metal looks promising for the use of hydrogen storage materials. This paper has the aim to study the hydriding properties of uranium alloys, including U-Ti U-Mo and U-Ni. The uranium alloys specimens were prepared by melting the constituent elements by means of simultaneous measurements of thermo-gravimetric and differential thermal analyses (TGA-DTA) and studied in as cast condition as hydrogen storage materials. Then samples were thermally treated under constant flow of hydrogen, at various temperatures between 573-973 0 K. The structural and absorption properties of the products obtained were examined by thermo-gravimetric analysis (TG), X-ray diffraction (XRD) and scanning electron microscopy (SEM). They slowly reacted with hydrogen to form the ternary hydride and the hydrogenated samples mainly consisted of the pursued ternary hydride bat contained also U or UO 2 and some transient phase. (authors)

Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

Energy Technology Data Exchange (ETDEWEB)

A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

2009-08-01

A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

Dissolution of metallic uranium and its alloys. Part 1. Review of analytical and process-scale metallic uranium dissolution

International Nuclear Information System (INIS)

Laue, C.A.; Gates-Anderson, D.; Fitch, T.E.

2004-01-01

This review focuses on dissolution/reaction systems capable of treating uranium metal waste to remove its pyrophoric properties. The primary emphasis is the review of literature describing analytical and production-scale dissolution methods applied to either uranium metal or uranium alloys. A brief summary of uranium's corrosion behavior is included since the corrosion resistance of metals and alloys affects their dissolution behavior. Based on this review, dissolution systems were recommended for subsequent screening studies designed to identify the best system to treat depleted uranium metal wastes at Lawrence Livermore National Laboratory (LLNL). (author)

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

Energy Technology Data Exchange (ETDEWEB)

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth; Arey, Bruce; Jana, Saumyadeep; Lavender, Curt; Joshi, Vineet

2018-06-01

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 grain 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.

Positron lifetime study of neutron-irradiated molybdenum

International Nuclear Information System (INIS)

Hinode, Kenji; Tanigawa, Shoichiro; Kumakura, Hiroaki; Doyama, Masao; Shiraishi, Kensuke.

1978-01-01

Annealing behavior of fast-neutron-irradiated molybdenum was studied by means of positron lifetime technique. It was found that Stage III annealing can be mainly identified as the vacancy migration process from the detailed analyses of data. The void growth after successive high temperature annealings was clearly detected through the changes of positron lifetime parameters. An attempt to analyse the size distribution of voids from positron lifetime spectra was presented, and discussions on the evaluation of void concentration from positron data are also given. (author)

Studies on neutron irradiation effects of iron alloys and nickel-base heat resistant alloys

International Nuclear Information System (INIS)

Watanabe, Katsutoshi

1987-09-01

The present paper describes the results of neutron irradiation effects on iron alloys and nickel-base heat resistant alloys. As for the iron alloys, irradiation hardening and embrittlement were investigated using internal friction measurement, electron microscopy and tensile testings. The role of alloying elements was also investigated to understand the irradiation behavior of iron alloys. The essential factors affecting irradiation hardening and embrittlement were thus clarified. On the other hand, postirradiation tensile and creep properties were measured of Hastelloy X alloy. Irradiation behavior at elevated temperatures is discussed. (author)

Void formation by annealing of neutron-irradiated plastically deformed molybdenum

International Nuclear Information System (INIS)

Petersen, K.; Nielsen, B.; Thrane, N.

1976-01-01

The positron annihilation technique has been used in order to study the influence of plastic deformation on the formation and growth of voids in neutron irradiated molybdenum single crystals treated by isochronal annealing. Samples were prepared in three ways: deformed 12-19% before irradiation, deformed 12-19% after irradiation, and - for reference purposes -non-deformed. In addition a polycrystalline sample was prepared in order to study the influence of the grain boundaries. All samples were irradiated at 60 0 C with a flux of 2.5 x 10 18 fast neutrons/cm 2 . After irradiation the samples were subjected to isochronal annealing. It was found that deformation before irradiation probably enhanced the formation of voids slightly. Deformation after irradiation strongly reduced the void formation. The presence of grain boundaries in the polycrystalline sample had a reducing influence on the growth of voids. (author)

Development of casting techniques for uranium and uranium alloys

International Nuclear Information System (INIS)

Singh, S.P.

2003-01-01

The casting process concerning furnace set-up, mould temperatures, pouring temperatures, out gassing, post heating, casting recovery and crucible and mould clean-up is discussed. Some applications of casting theory can be made in practice, but experience in handling the metal is most valuable in the successful solution of a new problem. The casting of uranium alloys using induction stirring of the melt to promote homogeneity in the casting is described. A few remarks are made concerning safety aspects associated with the casting of uranium

Development of metal uranium fuel and testing of construction materials (I-VI); Part I

International Nuclear Information System (INIS)

Mihajlovic, A.

1965-11-01

This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors

Theoretical study of the localization-delocalization transition in amorphous molybdenum-germanium alloys

International Nuclear Information System (INIS)

Ding, K.; Andersen, H.C.

1987-01-01

Electronic structure calculations were performed for amorphous germanium and amorphous alloys of molybdenum and germanium. The calculations used Harrison's universal linear-combination-of-atomic-orbitals parameters to generate one-electron Hamiltonians for structural configurations obtained from molecular-dynamics simulations. The density of states calculated for a model of a-Ge showed a distinct pseudogap, although with an appreciable density of states at the minimum. The states in the pseudogap are localized. As the concentration of Mo atoms increases, the pseudogap of the density of states is gradually filled up. The density of states at the Fermi energy calculated for our model of the alloys agrees quite well with that experimentally determined by Yoshizumi, Geballe, and co-workers. The localization index for the states at the Fermi energy is a decreasing function of Mo concentration in the range of 2--14 at. % Mo and the localization length is an increasing function of molybdenum concentration. These results are consistent with the experimental observation of an insulator-metal transition at about 10 at. % Mo

Preliminary Accident Analyses for Conversion of the Massachusetts Institute of Technology Reactor (MITR) from Highly Enriched to Low Enriched Uranium

Energy Technology Data Exchange (ETDEWEB)

Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, Erik H. [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Kaichao S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Newton, Jr., Thomas H. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

2013-09-30

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. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its 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 MITR. This report presents the preliminary accident analyses for MITR cores fueled with LEU monolithic U-Mo alloy fuel with 10 wt% Mo. Preliminary results demonstrate adequate performance, including thermal margin to expected safety limits, for the LEU accident scenarios analyzed.

Karakteristik kawat TMA (titanium molybdenum alloy) dan penggunaannya dalam perawatan ortodonti

OpenAIRE

Arifiani, Putri; Erwin Siregar, Erwin Siregar

2016-01-01

Kawat merupakan salah satu piranti yang penting dalam perawatan ortodonsia. Perkembangan terkini dari kawat ortodonsia menghasilkan beberapa jenis kawat dengan karakteristik yang berbeda-beda. Studi pustaka membahas karakteristik kawat ortodonsi beta titanium atau Titanium Molybdenum Alloy (TMA) dan penggunaannya dalam perawatan ortodonsi. Perbedaan karakteristik tiap kawat menjadi hal yang perlu dipertimbangkan secara klinis. Kawat beta titanium atau sering disebut juga dengan kawat TMA (Tit...

The nature of planar faults in a dilute molybdenum-boron alloy

International Nuclear Information System (INIS)

Chervinskii, V.I.; Kantor, M.M.; Novikov, I.I.; Sofronova, R.M.

1982-01-01

Planar faults on (100) planes in dilute molybdenum-boron alloys consist of a mono- or a bilayer of boron atoms. The displacement vectors are of the general type and for mono- and bilayer faults, respectively, where the component d is close to 1/6 and normal to the fault plane. The planar faults are probably an intermediate stage of MoB or Mo 2 BC growth. (author)

Influence of iron and beryllium additions on heat resistance of silicide coatings on TsMB-30 molybdenum alloy

International Nuclear Information System (INIS)

Zajtseva, A.L.; Fedorchuk, N.M.; Lazarev, Eh.M.; Korotkov, N.A.

1985-01-01

Alloying of titanium modified silicide coatings on TsMB-30 molybdenum alloy with iron or beryllium is stated to improve their protective properties. Coatings with low content of alloying elements have the best protective properties. Service life of coatings is determined by the formed oxide film and phase transformations taking place in the coating

A comparison of corrosion resistance of cobalt-chromium-molybdenum metal ceramic alloy fabricated with selective laser melting and traditional processing.

Science.gov (United States)

Zeng, Li; Xiang, Nan; Wei, Bin

2014-11-01

A cobalt-chromium-molybdenum alloy fabricated by selective laser melting is a promising material; however, there are concerns about the change in its corrosion behavior. The purpose of this study was to evaluate the changes in corrosion behavior of a cobalt-chromium-molybdenum alloy fabricated by the selective laser melting technique before and after ceramic firing, with traditional processing of cobalt-chromium-molybdenum alloy serving as a control. Two groups of specimens were designated as group selective laser melting and group traditional. For each group, 20 specimens with a cylindrical shape were prepared and divided into 4 cells: selective laser melting as-cast, selective laser melting fired in pH 5.0 and 2.5, traditional as-cast, and traditional fired in pH 5.0 and 2.5. Specimens were prepared with a selective laser melting system for a selective laser melting alloy and the conventional lost wax technique for traditional cast alloy. After all specimen surfaces had been wet ground with silicon carbide paper (1200 grit), each group of 10 specimens was put through a series of ceramic firing cycles. Microstructure, Vickers microhardness, surface composition, oxide film thickness, and corrosion behavior were examined for specimens before and after ceramic firing. Three-way ANOVA was used to evaluate the effect of porcelain firing and pH values on the corrosion behavior of the 2 alloys (α=.05). Student t tests were used to compare the Vickers hardness. Although porcelain firing changed the microstructure, microhardness, and x-ray photoelectron spectroscopy results, it showed no significant influence on the corrosion behavior of the selective laser melting alloy and traditional cast alloy (P>.05). No statistically significant influence was found on the corrosion behavior of the 2 alloys in different pH value solutions (P>.05). The porcelain firing process had no significant influence on the corrosion resistance results of the 2 alloys. Compared with traditional

Some properties of aluminum-uranium alloys in the cast, rolled and annealed conditions

International Nuclear Information System (INIS)

Jones, T.I.; McGee, I.J.; Norlock, L.R.

1960-06-01

The metallographic and hardness changes associated with the rolling and subsequent. annealing of aluminum alloys containing up to 30-wt.% uranium have been described. The alloys possessed good rolling properties. However the richer alloys were unusual in that after an initial reduction,, further cold rolling caused softening. In the alloy range examined, increasing uranium contents caused reduced preferred orientation. Qualitative explanations have been proposed to account for the observations on roll softening and preferred orientation. Heat-treating and ageing experiments confirmed that the solid solubility of uranium in aluminum is negligible. (author)

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

Science.gov (United States)

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.

Effect of neutron irradiation on vanadium alloys

International Nuclear Information System (INIS)

Braski, D.N.

1986-01-01

Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600 0 C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520 0 C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys

Effect of neutron irradiation on vanadium alloys

Energy Technology Data Exchange (ETDEWEB)

Braski, D.N.

1986-01-01

Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

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

Science.gov (United States)

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

2001-04-01

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

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

Energy Technology Data Exchange (ETDEWEB)

Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

2009-03-01

This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

A chemical approach toward low temperature alloying of immiscible iron and molybdenum metals

Energy Technology Data Exchange (ETDEWEB)

Nazir, Rabia [Department of Chemistry, Quaid-i-Azam University, Islamabad 45320 (Pakistan); Applied Chemistry Research Centre, Pakistan Council of Scientific and Industrial Research Laboratories Complex, Lahore 54600 (Pakistan); Ahmed, Sohail [Department of Chemistry, Quaid-i-Azam University, Islamabad 45320 (Pakistan); Mazhar, Muhammad, E-mail: mazhar42pk@yahoo.com [Department of Chemistry, University of Malaya, Lembah Pantai, 50603 Kuala Lumpur (Malaysia); Akhtar, Muhammad Javed; Siddique, Muhammad [Physics Division, PINSTECH, P.O. Nilore, Islamabad (Pakistan); Khan, Nawazish Ali [Material Science Laboratory, Department of Physics, Quaid-i-Azam University, Islamabad 45320 (Pakistan); Shah, Muhammad Raza [HEJ Research Institute of Chemistry, University of Karachi, Karachi 75270 (Pakistan); Nadeem, Muhammad [Physics Division, PINSTECH, P.O. Nilore, Islamabad (Pakistan)

2013-11-15

Graphical abstract: - Highlights: • Low temperature pyrolysis of [Fe(bipy){sub 3}]Cl{sub 2} and [Mo(bipy)Cl{sub 4}] homogeneous powder. • Easy low temperature alloying of immiscible metals like Fe and Mo. • Uniform sized Fe–Mo nanoalloy with particle size of 48–68 nm. • Characterization by EDXRF, AFM, XRPD, magnetometery, {sup 57}Fe Mössbauer and impedance. • Alloy behaves as almost superparamagnetic obeying simple –R(CPE)– circuit. - Abstract: The present research is based on a low temperature operated feasible method for the synthesis of immiscible iron and molybdenum metals’ nanoalloy for technological applications. The nanoalloy has been synthesized by pyrolysis of homogeneous powder precipitated, from a common solvent, of the two complexes, trisbipyridineiron(II)chloride, [Fe(bipy){sub 3}]Cl{sub 2}, and bipyridinemolybedenum(IV) chloride, [Mo(bipy)Cl{sub 4}], followed by heating at 500 °C in an inert atmosphere of flowing argon gas. The resulting nanoalloy has been characterized by using EDXRF, AFM, XRD, magnetometery, {sup 57}Fe Mössbauer and impedance spectroscopies. These results showed that under provided experimental conditions iron and molybdenum metals, with known miscibility barrier, alloy together to give (1:1) single phase material having particle size in the range of 48–66 nm. The magnetism of iron is considerably reduced after alloy formation and shows its trend toward superparamagnetism. The designed chemical synthetic procedure is equally feasible for the fabrication of other immiscible metals.

A chemical approach toward low temperature alloying of immiscible iron and molybdenum metals

International Nuclear Information System (INIS)

Nazir, Rabia; Ahmed, Sohail; Mazhar, Muhammad; Akhtar, Muhammad Javed; Siddique, Muhammad; Khan, Nawazish Ali; Shah, Muhammad Raza; Nadeem, Muhammad

2013-01-01

Graphical abstract: - Highlights: • Low temperature pyrolysis of [Fe(bipy) 3 ]Cl 2 and [Mo(bipy)Cl 4 ] homogeneous powder. • Easy low temperature alloying of immiscible metals like Fe and Mo. • Uniform sized Fe–Mo nanoalloy with particle size of 48–68 nm. • Characterization by EDXRF, AFM, XRPD, magnetometery, 57 Fe Mössbauer and impedance. • Alloy behaves as almost superparamagnetic obeying simple –R(CPE)– circuit. - Abstract: The present research is based on a low temperature operated feasible method for the synthesis of immiscible iron and molybdenum metals’ nanoalloy for technological applications. The nanoalloy has been synthesized by pyrolysis of homogeneous powder precipitated, from a common solvent, of the two complexes, trisbipyridineiron(II)chloride, [Fe(bipy) 3 ]Cl 2 , and bipyridinemolybedenum(IV) chloride, [Mo(bipy)Cl 4 ], followed by heating at 500 °C in an inert atmosphere of flowing argon gas. The resulting nanoalloy has been characterized by using EDXRF, AFM, XRD, magnetometery, 57 Fe Mössbauer and impedance spectroscopies. These results showed that under provided experimental conditions iron and molybdenum metals, with known miscibility barrier, alloy together to give (1:1) single phase material having particle size in the range of 48–66 nm. The magnetism of iron is considerably reduced after alloy formation and shows its trend toward superparamagnetism. The designed chemical synthetic procedure is equally feasible for the fabrication of other immiscible metals

Amorphous uranium alloy and use thereof

International Nuclear Information System (INIS)

Gambino, R.J.; McElfresh, M.W.; McGuire, T.R.; Plaskett, T.S.

1991-01-01

An amorphous alloy containing uranium and a member selected from the group N, P, As, Sb, Bi, S, Se, Te, Po and mixtures thereof; and use thereof for storage medium, light modulator or optical isolator. (author) figs

Equations of state for enriched uranium and uranium alloy to 3500 MPa

International Nuclear Information System (INIS)

Bai Chaomao; Hai Yuying; Liu Jenlong; Li Zhenrong

1990-04-01

The volume compressions of 6 kinds of cast materials including enriched uranium, poor uranium, U-0.57 wt% Ti, U-0.33 wt% Nb, U-2.85 wt% Nb and U-7.5 wt% Nb-3.3 wt% Zr have been determined by monitoring piston displacements in a piston cylinder apparatus with double strengthening rings to 3500 MPa at room temperature. The dilation of the cylinder vessel and the press deformation were corrected by some experiments. The calculational data free from using the standard sample closed with used standard sample. The volume compressions of enriched uranium and poor uranium are nearly coincident. Pure uranium is more compressible than uranium alloys. These values of enriched uranium are in close agreement with values of Bridgman's pure uranium. The fitting coefficients of Bridgman's polynomial and Anderson's equation of state and isothermal bulk modules for the above materials are given

The behaviour under irradiation of molybdenum matrix for inert matrix fuel containing americium oxide (CerMet concept)

Energy Technology Data Exchange (ETDEWEB)

D' Agata, E., E-mail: elio.dagata@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); Knol, S.; Fedorov, A.V. [Nuclear Research and Consultancy Group, P.O. Box 25, 1755 ZG Petten (Netherlands); Fernandez, A.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Klaassen, F. [Nuclear Research and Consultancy Group, P.O. Box 25, 1755 ZG Petten (Netherlands)

2015-10-15

Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors or Accelerator Driven System (ADS, subcritical reactors dedicated to transmutation) of long-lived nuclides like {sup 241}Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. In order to safely burn americium in a fast reactor or ADS, it must be incorporated in a matrix that could be metallic (CerMet target) or ceramic (CerCer target). One of the most promising matrix to incorporate Am is molybdenum. In order to address the issues (swelling, stability under irradiation, gas retention and release) of using Mo as matrix to transmute Am, two irradiation experiments have been conducted recently at the High Flux Reactor (HFR) in Petten (The Netherland) namely HELIOS and BODEX. The BODEX experiment is a separate effect test, where the molybdenum behaviour is studied without the presence of fission products using {sup 10}B to “produce” helium, the HELIOS experiment included a more representative fuel target with the presence of Am and fission product. This paper covers the results of Post Irradiation Examination (PIE) of the two irradiation experiments mentioned above where molybdenum behaviour has been deeply investigated as possible matrix to transmute americium (CerMet fuel target). The behaviour of molybdenum looks satisfying at operating temperature but at high temperature (above 1000 °C) more investigation should be performed.

Irradiation-assisted stress corrosion cracking in HTH Alloy X-750 and Alloy 625

International Nuclear Information System (INIS)

Bajaj, R.; Mills, W.J.; Lebo, M.R.; Hyatt, B.Z.; Burke, M.G.

1995-01-01

In-reactor testing of bolt-loaded compact tension specimens was performed in 360 C water to determine the irradiation-assisted stress corrosion cracking (IASCC) behavior of HTH Alloy X-750 and direct-aged Alloy 625. New data confirm previous results showing that high irradiation levels reduce SCC resistance in Alloy X-750. Heat-to-heat variability correlates with boron content, with low boron heats showing improved IASCC properties. Alloy 625 is resistant to IASCC, as no cracking was observed in any Alloy 625 specimens. Microstructural, microchemical and deformation studies were performed to characterize the mechanisms responsible for IASCC in Alloy X-750 and the lack of an effect in Alloy 625. The mechanisms under investigation are: boron transmutation effects, radiation-induced changes in microstructure and deformation characteristics, and radiation-induced segregation. Irradiation of Alloy X-750 caused significant strengthening and ductility loss that was associated with the formation of cavities and dislocation loops. High irradiation levels did not cause significant segregation of alloying or trace elements in Alloy X-750. Irradiation of Alloy 625 resulted in the formation of small dislocation loops and a fine body-centered-orthorhombic phase. The strengthening due to the loops and precipitates was apparently offset by a partial dissolution of γ double-prime precipitates, as Alloy 625 showed no irradiation-induced strengthening or ductility loss. In the nonirradiated condition, an IASCC susceptible HTH heat containing 28 ppm B showed grain boundary segregation of boron, whereas a nonsusceptible HTH heat containing 2 ppm B and Alloy 625 with 20 ppm B did not show significant boron segregation. Transmutation of boron to helium at grain boundaries, coupled with matrix strengthening, is believed to be responsible for IASCC in Alloy X-750, and the absence of these two effects results in the superior IASCC resistance displayed by Alloy 625

The effect of strain rate and temperature on the tensile behaviour of uranium 2 w/o molybdenum

International Nuclear Information System (INIS)

Harding, J.; Boyd, G.A.C.

1983-01-01

This report describes the uniaxial tensile behaviour of uranium 2 w/o molybdenum alloy over a wide range of temperature and strain rate. Specimen blanks taken from co-reduced and extruded U 2 w/o Mo rods were given one of two heat treatments. Longitudinal tensile test pieces, taken from these blanks at near surface locations were tested in the temperature range -150 deg C to +100 deg C at strain rates from quasistatic (10 -4 s -1 ) to 10 3 s -1 . To achieve this range of testing rates three machines were required: an Instron screw driven machine for rates up to 0.1 s -1 , a second specially constructed hydraulic machine for the range 0.1 s -1 to 50 s -1 and a drop weight machine for the highest strain rates. The ways in which the mechanical properties - elongation to fracture, flow stresses and ultimate tensile stress - vary with both temperature and strain rate are presented and discussed for material in both heat treatment conditions. (author)

Evaluation of magnetite nanoparticles as molybdenum ions adsorbent

International Nuclear Information System (INIS)

Holland, Helber; Yamaura, Mitiko; Sousa, Jose Silva; Freitas, Antonio Alves

2011-01-01

Molybdenum-99 is the generator radionuclide of the most used radioisotope for preparation of radiopharmaceuticals with diagnostic purposes in nuclear medicine, technetium-99m (Tc-99m). One way of Mo-99 obtaining is as fission product of irradiated uranium targets in reactor. In this work, the potential application of magnetite particles in the separation of Mo-99 from a dissolution solution of U targets was evaluated. Synthetic magnetite nanoparticles were prepared by alkaline precipitation method from Fe 2+ ions and heat-treated via microwave irradiation in a conventional household oven. Adsorption kinetics was studied. It was observed that the adsorption of Mo by magnetite nanoparticles is fast and followed the model of pseudo-second order. (author)

Fundamental irradiation studies on vanadium alloys

International Nuclear Information System (INIS)

Loomis, B.A.; Garner, F.A.; Ermi, A.M.

1985-05-01

A joint experiment on the irradiation response of simple vanadium alloys has been initiated under the auspices of the DAFS and BES progams. Specimen fabrication is nearly complete and the alloys are expected to be irradiated in lithium in FFTF-MOTA Cycles 7 and 8

Study of internal oxidation kinetics of molybdenum base alloys

International Nuclear Information System (INIS)

Krushinskij, Yu.Yu.; Belyakov, B.G.; Belomyttsev, M.Yu.

1989-01-01

Metallographic and microdurometric method as well as new technique were used to study kinetics of internal oxidation (IO). It is shown that study of IO kinetics on the base of metallographic measurements of layers depth is not correct because it is related with insufficient sensitivity of the method. IO kinetics under conditions of formation of molybdenum oxide layer on saturated material surface as well as IO of alloy with high carbon content were investigated. Oxide film formation does not affect the IO kinetics; decarburization observed along with oxidation increases the apparent activation energy and K exponent on time dependence of diffusion layer depth

Study of the pyrophoric characteristics of uranium-iron alloys; Etude du caractere pyrophorique des alliages uranium fer

Energy Technology Data Exchange (ETDEWEB)

Duplessis, X

2000-02-23

The objective of the study is to understand the pyrophoric characteristics of uranium-iron alloys. In order to carry out this research we have elected to use uranium-iron alloy powder with granules of 200 {mu}m and 1000 {mu}m diameter with 4%, 10.8% and 14% iron content. The experiments were performed on small samples of few milligrams and on larger quantities of few hundred grams. The main conclusions obtained are the followings: -The reaction start at 453 K (180 deg. C) and the ignition at 543 K (270 deg. C) - The influence of the specific area seems more important than the iron concentration in the alloys - When the alloy ignites, the fire spreads quickly and the alloy rapidly consumes. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Devaraj, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Prabhakaran, Ramprashad [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Hu, Shenyang Y. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); McGarrah, Eric J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States). Environmental Molecular Sciences Lab. (EMSL)

2016-04-12

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

Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

International Nuclear Information System (INIS)

Loelgen, R.H.; Cundy, M.R.; Schuele, W.

1977-01-01

A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this obervation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. Measurements of radiation enhanced diffusion are less time consuming and expensive than irradiation creep tests and information on this property can be obtained rather quickly, prior to the selection of stainless steel alloys for creep tests. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. Finally, a few uniaxial tensile creep tests will be performed in fully instrumented rigs. (Auth.)

Irradiation creep in simple binary alloys

International Nuclear Information System (INIS)

Nagakawa, J.; Sethi, V.K.; Turner, A.P.L.

1981-07-01

Creep enhancement during 21-MeV deuteron irradiation was examined at 350 0 C for two simple binary alloys with representative microstructures, i.e., solid-solution (Ni - 4 at. % Si) and precipitation-hardened (Ni - 12.8 at. % Al) alloys. Coherent precipitates were found to be very effective in suppressing irradiation-enhanced creep. Si solute atoms depressed irradiation creep moderately and caused irradiation hardening via radiation-induced segregation. The stress-dependence of irradiation creep in Ni - 4 at. % Si should a transition, which seems to reflect a change of mechanism from dislocation climb due to stress-induced preferential absorption (SIPA) to climb-controlled dislocation glide enhanced by irradiation

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

Energy Technology Data Exchange (ETDEWEB)

Coffey, Greg W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Meinhardt, Kerry D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pederson, Larry R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-03-01

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

Estimation of the effect of molybdenum on chemical and electrochemical stability of iron-based alloys

International Nuclear Information System (INIS)

Tyurin, A.G.

2003-01-01

The E-pH diagram for Mo-H 2 O system is made more precise. It is shown that a passivating film on molybdenum in weakly acid, neutral and alkali solutions may constitute MoO 2 only. In strongly acid solutions at anodic polarization the film should transform according to the following scheme: MoO 2 → Mo 4 O 11 → Mo 9 O 26 → MoO 3 . Sections of a Fe-Mo-O system phase diagram and a Fe-Mo-H 2 O system E-pH diagram at 25 deg C are plotted. MoO 2 is found to be a product of iron-molybdenum alloy oxidation in the air and in water. For the system of alloy Kh17N13M2-H 2 O the section of a E-pH diagram is plotted at 25 deg C [ru

Analytic chemistry of molybdenum

International Nuclear Information System (INIS)

Parker, G.A.

1983-01-01

Electrochemical, colorimetric, gravimetric, spectroscopic, and radiochemical methods for the determination of molybdenum are summarized in this book. Some laboratory procedures are described in detail while literature citations are given for others. The reader is also referred to older comprehensive reviews of the analytical chemistry of molybdenum. Contents, abridged: Gravimetric methods. Titrimetric methods. Colorimetric methods. X-ray fluorescence. Voltammetry. Catalytic methods. Molybdenum in non-ferrous alloys. Molydbenum compounds

Irradiation creep of dispersion strengthened copper alloy

Energy Technology Data Exchange (ETDEWEB)

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A. [and others

1997-04-01

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

Irradiation creep of dispersion strengthened copper alloy

International Nuclear Information System (INIS)

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

1997-01-01

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

Irradiation growth in zirconium alloys: a review

International Nuclear Information System (INIS)

Fidleris, V.

1980-09-01

The change in shape during irradiation without external stress, irradiation growth, was first discovered in uranium and later in graphite, zirconium and other core materials which exhibit anisotropic physical properties. The direction of maximum growth of metals invariably corresponds with the direction of minimum thermal expansion. In polycrystalline zirconium alloys growth is positive in the direction of maximum deformation during fabrication and in other directions it can be either positive or negative depending on the preferred orientation of grains (crystallographic texture). Growth increases gradually with temperature between 300 K and 620 K and rapidly with fluence up to about 1 x 10 25 n.m. -2 (Eμ1 MeV). At higher fluences the growth appears to saturate in annealed materials and reach a steady rate approximately proportional to dislocation density in cold-worked materials. Above 600 K both annealed and cold-worked materials have similar steady growth rates. Irradiation growth is caused by the segregation to different sinks of the vacancies and interstitials generated by irradiation, but the dominant types of sinks for each type of point defect and the mode of transport of the point defects to sinks cannot therefore be predicted theoretically. For the purpose of designing reactor core components empirical equations have been derived that can satisfactorily predict the steady state growth behaviour from texture and microstructure. (auth)

Swelling in neutron-irradiated titanium alloys

International Nuclear Information System (INIS)

Peterson, D.T.

1982-04-01

Immersion density measurements have been performed on a series of titanium alloys irradiated in EBR-II to a fluence of 5 x 10 22 n/cm 2 (E > 0.1 MeV) at 450 and 550 0 C. The materials irradiated were the near-alpha alloys Ti-6242S and Ti-5621S, the alpha-beta alloy Ti-64, and the beta alloy Ti-38644. Swelling was observed in all alloys with the greater swelling being observed at 550 0 C. Microstructural examination revealed the presence of voids in all alloys. Ti-38644 was found to be the most radiation resistant. Ti-6242S and Ti-5621S also displayed good radiation resistance, whereas considerable swelling and precipitation were observed in Ti-64 at 550 0 C

Crystallographic study of the tempering by irradiation of cold-worked uranium (1960); Etude cristallographique du revenu de l'uranium ecroui par irradiation (1960)

Energy Technology Data Exchange (ETDEWEB)

Tardivon, D [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

We have studied the phenomenon of the tempering of cold-worked uranium under the action of irradiation by observing the narrowing of the (114) and (133) X-ray diffraction lines as a function of the irradiation level. Simultaneously we have studied the broadening of the 114 line of a recrystallised uranium as a function of the irradiation level. The irradiation temperature was always less than 60 deg. C. Of these two processes, the first is the fastest. We have observed a saturation of the irradiation tempering for a flux of 10{sup 18} n/cm{sup 2}; we deduce from this the dimensions of the volume perturbed by one fission atom to be 10{sup -17} cm{sup 3}. (author) [French] Nous avons etudie le phenomene de revenu par irradiation d'echantillons d'uranium ecroui, en observant l'affinement des raies de diffraction de rayons X (114) et (133) en fonction du taux d'irradiation. Parallelement nous avons etudie l'elargissement de la raie (114 ) d'un uranium recristallise en fonction du taux d'irradiation. La temperature d'irradiation est toujours restee inferieure a 60 deg. C. De ces deux processus le premier est le plus rapide. Nous avons observe une saturation du revenu par irradiation pour un flux de 10{sup 18} n/cm{sup 2}; on en deduit une valeur du volume de la perturbation creee par un atome de fission egale a 10{sup -17} cm{sup 3}. (auteur)

Mechanical properties of molybdenum alloyed liquid phase-sintered tungsten-based composites

International Nuclear Information System (INIS)

Kemp, P.B.; German, R.M.

1995-01-01

Tungsten-based composites are fabricated from mixed elemental powders using liquid phase sintering, usually with a nickel-iron matrix. During sintering, the tungsten undergoes grain growth, leading to microstructure coarsening that lowers strength but increases ductility. Often the desire is to increase strength at the sacrifice of ductility, and historically, this has been performed by postsintering deformation. There has been considerable research on alloying to adjust the as-sintered mechanical properties to match those of swaged alloys. Prior reports cover many additions, seemingly including much of the periodic table. Unfortunately, many of the modified alloys proved disappointing, largely due to degraded strength at the tungsten-matrix interface. Of these modified alloys, the molybdenum-containing systems exhibit a promising combination of properties, cost, and processing ease. For example, the 82W-8Mo-7Ni-3Fe alloy gives a yield strength that is 34% higher than the equivalent 90W-7Ni-3Fe alloy (from 535 to 715 MPa) but with a 33% decrease in fracture elongation (from 30 to 20% elongation). This article reports on experiments geared to promoting improved properties in the W-Mo-Ni-Fe alloys. However, unlike the prior research which maintained a constant Ni + Fe content and varied the W:Mo ratio, this study considers the Mo:(Ni + Fe) ratio effect for 82, 90, and 93 wt pct W

Liquid phase sintering of carbides using a nickel-molybdenum alloy

International Nuclear Information System (INIS)

Barranco, J.M.; Warenchak, R.A.

1987-01-01

Liquid phase vacuum sintering was used to densify four carbide groups. These were titanium carbide, tungsten carbide, vanadium carbide, and zirconium carbide. The liquid phase consisted of nickel with additions of molybdenum of from 6.25 to 50.0 weight percent at doubling increments. The liquid phase or binder comprised 10, 20, and 40 percent by weight of the pressed powders. The specimens were tested using 3 point bending. Tungsten carbide showed the greatest improvement in bend rupture strength, flexural modulus, fracture energy and hardness using 20 percent binder with lesser amounts of molybdenum (6.25 or 12.5 wt %) added to nickel compared to pure nickel. A refinement in the carbide microstructure and/or a reduction in porosity was seen for both the titanium and tungsten carbides when the alloy binder was used compared to using the nickel alone. Curves depicting the above properties are shown for increasing amounts of molybdenum in nickel for each carbide examined. Loss of binder phase due to evaporation was experienced during heating in vacuum at sintering temperatures. In an effort to reduce porosity, identical specimens were HIP processed at 15 ksi and temperatures averaging 110 C below the sintering g temperature. The tungsten carbide and titanium carbide series containing 80 and 90 weight percent carbide phase respectively showed improvement properties after HIP while properties decreased for most other compositions

Uranium briquettes for irradiation target

Energy Technology Data Exchange (ETDEWEB)

Saliba-Silva, Adonis Marcelo; Garcia, Rafael Henrique Lazzari; Martins, Ilson Carlos; Carvalho, Elita Fontenele Urano de; Durazzo, Michelangelo, E-mail: saliba@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

Direct irradiation on targets inside nuclear research or multiple purpose reactors is a common route to produce {sup 99}Mo-{sup 99m}Tc radioisotopes. Nevertheless, since the imposed limits to use LEU uranium to prevent nuclear armament production, the amount of uranium loaded in target meats has physically increased and new processes have been proposed for production. Routes using metallic uranium thin film and UAl{sub x} dispersion have been used for this purpose. Both routes have their own issues, either by bringing difficulties to disassemble the aluminum case inside hot cells or by generating great amount of alkaline radioactive liquid rejects. A potential route might be the dispersion of powders of LEU metallic uranium and nickel, which are pressed as a blend inside a die and followed by pulse electroplating of nickel. The electroplating provides more strength to the briquettes and creates a barrier for gas evolution during neutronic disintegration of {sup 235}U. A target briquette platted with nickel encapsulated in an aluminum case to be irradiated may be an alternative possibility to replace other proposed targets. This work uses pulse Ni-electroplating over iron powder briquette to simulate the covering of uranium by nickel. The following parameters were applied 10 times for each sample: 900Hz, -0.84A/square centimeters with duty cycle of 0.1 in Watts Bath. It also presented the optical microscopy analysis of plated microstructure section. (author)

Uranium briquettes for irradiation target

International Nuclear Information System (INIS)

Saliba-Silva, Adonis Marcelo; Garcia, Rafael Henrique Lazzari; Martins, Ilson Carlos; Carvalho, Elita Fontenele Urano de; Durazzo, Michelangelo

2011-01-01

Direct irradiation on targets inside nuclear research or multiple purpose reactors is a common route to produce 99 Mo- 99m Tc radioisotopes. Nevertheless, since the imposed limits to use LEU uranium to prevent nuclear armament production, the amount of uranium loaded in target meats has physically increased and new processes have been proposed for production. Routes using metallic uranium thin film and UAl x dispersion have been used for this purpose. Both routes have their own issues, either by bringing difficulties to disassemble the aluminum case inside hot cells or by generating great amount of alkaline radioactive liquid rejects. A potential route might be the dispersion of powders of LEU metallic uranium and nickel, which are pressed as a blend inside a die and followed by pulse electroplating of nickel. The electroplating provides more strength to the briquettes and creates a barrier for gas evolution during neutronic disintegration of 235 U. A target briquette platted with nickel encapsulated in an aluminum case to be irradiated may be an alternative possibility to replace other proposed targets. This work uses pulse Ni-electroplating over iron powder briquette to simulate the covering of uranium by nickel. The following parameters were applied 10 times for each sample: 900Hz, -0.84A/square centimeters with duty cycle of 0.1 in Watts Bath. It also presented the optical microscopy analysis of plated microstructure section. (author)

Enhancement of uranium-accumulating ability of microorganisms by irradiation

Energy Technology Data Exchange (ETDEWEB)

Sakaguchi, Takashi; Nakajima, Akira; Tsuruta, Takehiko [Miyazaki Medical Coll., Kiyotake (Japan)

1998-01-01

Some microorganisms having excellent ability to accumulate uranium were isolated, from soil and water systems in and around the Ningyo-toge Station of Power Reactor and Nuclear Fuel Development Corporation. The enhancement of uranium-accumulating ability of microorganisms by electron-beam irradiation was examined, and the ability of JW-046 was increased 3-5% by the irradiation. The irradiation affect the growth of some of microorganisms tested. (author)

X-ray diffraction (XRD) characterization of microstrain in some iron and uranium alloys

International Nuclear Information System (INIS)

Kimmel, G.; Dayan, D.; Frank, G.A.; Landau, A.

1996-01-01

The high linear attenuation coefficient of steel, uranium and uranium based alloys is associated with the small penetration depth of X-rays with the usual wavelength used for diffraction. Nevertheless, by using the proper surface preparation technique, it is possible of obtaining surfaces with bulk properties (free of residual mechanical microstrain). Taking advantage of the feasibility to obtain well prepared surfaces, extensive work has been conducted in studying XRD line broadening effects from flat polycrystalline samples of steel, uranium and uranium alloys

Determination of molybdenum in uranium by differential pulse polarography

International Nuclear Information System (INIS)

Purwanto, A.; Iswani, G.S.

1996-01-01

The separation an determination of Mo (VI) in uranium dioxid (UO 2 ) samples by polarography method were studied. The determination of Mo(VI) was based on electrochemistry reduction of molybdenum (VI) quinolinolate complex (Mo VI O 2 Q 2 ). To decrease the matrix influence, this complex was extracted into chloroform. After chloroform phase was evaporate at room temperature, residues were dissolved with N,N-Dimethyl formamide (DMF). Potential wave reduction of Mo(VI) at -0,48 volt versus Ag/AgCl/KCl saturated in supporting electrolyte buffer acetate 1 M (1M CH 3 COOH-1M CH 3 COONH 4 ). Extraction was done in 0.3 M H 2 SO 4 media at optimum pH of 2.0 and 4 % oxine in 0.5 M H 2 SO 4 was used as complex compound. Extraction two times with 10 ml chloroform for 5 minutes each ratio of organic water was 1:5. This method was used to determine Mo (VI) in concentration range of 4.8 -12 μg in the presence of 200 mg uranium. It was found recovery of Mo (VI) was 95%. Mo (VI) contents in UO 2 samples determined by standard addition was 5.26±0.41 μg/g. (author)

The effect of strain rate and temperature on the tensile behaviour of uranium - 2sup(w)/o molybdenum

International Nuclear Information System (INIS)

Harding, J.; Boyd, G.A.C.

1983-01-01

This report describes the uniaxial tensile behaviour of uranium 2 w/o molybdenum alloy over a wide range of temperature and strain rate. Specimen blanks taken from co-reduced and extruded U2 w/o Mo rods were given one of two heat treatments. Longitudinal tensile test pieces, taken from these blanks at near surface locations were tested in the temperature range -150 deg C to +100 deg C at strain rates from quasistatic (10 -4 s -1 ) to 10 3 s -1 . To achieve this range of testing rates three machines were required: an Instron screw driven machine for rates up to 0.1 s -1 , a second specially constructed hydraulic machine for the range 0.1 s -1 to 50 s -1 and a drop weight machine for the highest strain rates. The ways in which the mechanical properties - elongation to fracture, flow stresses and ultimate tensile stress - vary with both temperature and strain rate are presented and discussed for material in both heat treatment conditions. (author)

Comparison of ion irradiation damage in three grades of unalloyed molybdenum

International Nuclear Information System (INIS)

Bradley, E.R.

1976-01-01

Transmission electron microscopy has been employed to characterize the ion irradiation damage in three grades of unalloyed molybdenum. The materials were irradiated with 5 MeV Ni 2+ ions at temperatures of 900 and 1000 0 C. Major differences exist in both the void and dislocation components of the damage and are attributed to differences in the carbon content of the three materials. A model, whereby carbon atoms segregate to small loops and decrease the bias for self interstitials, is used to qualitatively explain the observed variations in microstructure

Study on thermo-oxide layers of uranium-niobium alloy

International Nuclear Information System (INIS)

Luo Lizhu; Yang Jiangrong; Zhou Ping

2010-01-01

Surface oxides structure of uranium-niobium alloys which were annealed under different temperatures (room temperature, 100, 200, 300 degree C, respectively)in air were studied by X-ray photoelectron spectroscopy (XPS) analysis and depth profile. Thickness of thermo-oxide layers enhance with the increasing oxide temperature, and obvious changes to oxides structure are observed. Under different delt temperatures, Nb 2 O 5 are detected on the initial surface of U-Nb alloys, and a layer of NbO mixed with some NbO x (0 2 O 5 and Nb metal. Dealing samples in air from room temperature to 200 degree C, non-stoichiometric UO 2+x (UO 2 + interstitial oxygen, P-type semiconductor) are found on initial surface of U-Nb alloys, which has 0.7 eV shift to lower binding energy of U 4f 7/2 characteristics comparing to that of UO 2 . Under room temperature, UO 2 are commonly detected in the oxides layer, while under temperature of 100 and 200 degree C, some P-type UO 2+x are found in the oxide layers,which has a satellite at binding energy of 396.6 eV. When annealing at 300 degree C, higher valence oxides, such as U 3 O 8 or UO x (2 5/2 and U 4f 7/2 peaks are 392.2 and 381.8 eV, respectively. UO 2 mixed uranium metal are the main compositions in the oxide layers. From the results, influence of temperature to oxidation of uranium is more visible than to niobium in uranium-niobium alloys. (authors)

Recovery of molybdenum using alumina microspheres and precipitation with selective organic reagents

International Nuclear Information System (INIS)

Carvalho, Fatima Maria Sequeira de; Abrao, Alcidio

1998-01-01

In this paper is presented a study for the optimization of dissolution of the UAL x plates used for irradiation and production of radiomolybdenum. The alloy is dissolved in nitric acid with mercury as catalyst. The separation and concentration of the molybdenum was achieved using a chromatographic grade alumina microspheres column. the purified eluted molybdenum is finally precipitated using one of the selective reagents: alizarine blue, α,α'- bipyridine and 1,10-phenanthroline. Any one of the obtained precipitate can be fired to the molybdenum trioxide. The interference of the following elements was studied: Re(VII), U(VI), Cr(VI), W(VI), V(V), Te(IV), Ti(IV), Zr(IV), Th(IV), Fe(III), Au(III), Ru(III), Al(III), Bi(III), Sb(III), Ce(IV), Pr(III), Sc(III), Y(III), Sm(III), Ba(II), Sr(II), Ni(II), Co(II), Cs(I). The molybdenum precipitates were characterized by gravimetric, CHN, TG, DTG, IR and X-ray diffraction analyses. (author)

Determination of uranium in fissium-uranium alloy and in fissium dross

International Nuclear Information System (INIS)

Bodnar, L.Z.

1976-01-01

Dissolution and analysis techniques for fissium-uranium alloy and fissium dross are described. The fuming technique of dissolution effectively eliminated all interferring elements in the titration determination of U. The results from the semiquantitative analysis of fission dross by spark source mass spectrometry were tabulated

Neutronic analysis for the fission Mo99 production by irradiation of leu targets in TRIGA 14 MW reactor

International Nuclear Information System (INIS)

Dulugeac, S. D.; Mladin, M.; Budriman, A. G.

2013-01-01

Molybdenum production can be a solution for the future in the utilization of the Romanian TRIGA, taking into account the international market supply needs. Generally, two different techniques are available for Mo 99 production for use in medical Tc 99 generation.The first one is based on neutron irradiation of molybdenum targets of natural isotopic composition or enriched in Mo 98 . In a second process, Mo 99 is obtained as a result of the neutron induced fission of U 235 according to U 235 (n,f) Mo 99 . The objectives of the paper are related to Mo 99 production as a result of fission. Neutron physics parameters are determined and presented, such as: thermal flux axial distribution for the critical reactor at 10 MW inside the irradiation location; reactivity introduced by three Uranium foil containers; neutron fluxes and fission rates in the Uranium foils; released and deposited power in the Uranium foils; Mo 99 activity in the Uranium foils. (authors)

Transpassive dissolution of alloy 625, chromium, nickel, and molybdenum in high-temperature solutions containing hydrochloric acid and oxygen

International Nuclear Information System (INIS)

Kritzer, P.; Boukis, N.; Dinjus, E.

2000-01-01

Coupons of nickel, molybdenum, chromium, and the nickel-based Alloy 625 (UNS 06625) were corroded in strongly oxidizing hydrochloric acid (HCl) solutions at 350 C and a pressure (p) of 24 MPa, with reaction times between 0.75 h and 50 h. For Alloy 625, the effect of surface roughness also was investigated. Nickel and molybdenum showed strong material loss after only 5 h of reaction as a result of the instability of the solid oxides formed under experimental conditions. The attack on chromium started at the grain boundaries. At longer reaction times, thick, spalling oxide layers formed on the surface. The attack on Alloy 625 also started at the grain boundaries and at inclusions leading to the formation of small pits. On polished surfaces, the growth of these pits occurred faster than on nonpolished surfaces, but fewer pits grew. Corrosion products formed at the surface consisted of oxygen and chromium. On isolated spots, nickel- and chlorine-containing products also were found

Development of very-high-density low-enriched-uranium fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Trybus, C.L.; Wiencek, T.C.

1997-01-01

Following a hiatus of several years and following its successful development and qualification of 4.8 g U cm -3 U 3 Si 2 -Al dispersion fuel for application with low-enriched uranium in research and test reactors, the US Reduced Enrichment for Research and Test Reactors program has embarked on the development of even-higher-density fuels. Our goal is to achieve uranium densities of 8-9 g cm -3 in aluminum-based dispersion fuels. Achieving this goal will require the use of high-density, γ-stabilized uranium alloy powders in conjunction with the most-advanced fuel fabrication techniques. Key issues being addressed are the reaction of the fuel alloys with aluminum and the irradiation behavior of the fuel alloys and any reaction products. Test irradiations of candidate fuels in very-small (micro) plates are scheduled to begin in the Advanced Test Reactor during June, 1997. Initial results are expected to be available in early 1998. We are performing out-of-reactor studies on the phase structure of the candidate alloys on diffusion of the matrix material into the aluminum. In addition, we are modifying our current dispersion fuel irradiation behavior model to accommodate the new fuels. Several international partners are participating in various phases of this work. (orig.)

A spectroscopic study of uranium and molybdenum complexation within the pore channels of hybrid mesoporous silica

Energy Technology Data Exchange (ETDEWEB)

Charlot, Alexandre [CEA, DEN, DTDC, SPDE, Laboratoire des Procedes Supercritiques de Separation, Bagnols-sur-Ceze (France); CEA, DEN, DTDC, SPDE, Laboratoire de Developpement des Procedes de Separation, Bagnols-sur-Ceze (France); Dumas, Thomas [CEA, DEN, DTDC, SPDE, Laboratoire d' Interaction Ligands Actinides, Bagnols-sur-Ceze (France); Solari, Pier L. [Synchrotron SOLEIL, L' Orme des Merisiers, Saint-Aubin, Gif-sur-Yvette (France); Cuer, Frederic [CEA, DEN, DTDC, SPDE, Laboratoire de Developpement des Procedes de Separation, Bagnols-sur-Ceze (France); Grandjean, Agnes [CEA, DEN, DTDC, SPDE, Laboratoire des Procedes Supercritiques de Separation, Bagnols-sur-Ceze (France)

2017-01-18

To enable the reduction of the environmental impact of nuclear energy generation, in this paper, we link the molecular and macroscopic behaviour of a functionalized material (TR rate at SBA15) used to extract uranium from sulfuric media. Two organic 3-[N,N-di(2-ethylhexyl)carbamoyl]-3-[ethoxy(hydroxy)phosphoryl]propanoic acid (TR) molecules grafted onto the solid are involved in the extraction process and form a 2:1 TR-U complex. FTIR and extended X-ray absorption fine structure (EXAFS) spectroscopic analyses show that the TR-U bond is realized through a phosphonate group in a monodentate fashion below pH 3, in agreement with the macroscopic observations. The first coordination sphere of the uranyl ion is completed by two monodentate sulfate ions and one water molecule. Above pH 3, the TR contribution decreases, and other inner-sphere complexes appear, which is consistent with the increased extraction observed on the macroscopic scale. Molybdenum, a competitor element, reduces the uranium extraction capacity but not its speciation, whereas polyoxomolybdates form inside the pores from the molybdenum in solution. (copyright 2017 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Effect of high concentration of molybdenum on the structure and properties of niobium

International Nuclear Information System (INIS)

Trekina, M.I.; Vasil'eva, E.V.

1989-01-01

The effect of alloying of 20,25 and 30 % Mo on the structure and mechanical properties of niobium is studied. It is shown that niobium alloying with molybdenum in the studied concentration leads to grain grinding, which increases with the molybdenum content growth in the alloy. The effective energy values of recrystallization activation of the studied niobium and molybdenum alloys are determined. The high hardness level at some plasticity and deformability of niobium alloy with 20 % Mo is established

Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

Directory of Open Access Journals (Sweden)

Jaroszewicz Janusz

2014-07-01

Full Text Available The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

Kinetic parameters of nitridation of molybdenum and niobium alloys with various structure states

International Nuclear Information System (INIS)

Solodkin, G.A.; Bulgach, A.A.; Likhacheva, T.E.

1985-01-01

Effect of preliminary plastic strain under rolling on kinetic parameters of nitridation of VN-2AEh, VN-3 niobium alloys and molybdenum alloy with hafnium is investigated. Extreme character of dependence of kinetic parameters of nitridation on the degree of reduction under rolling is determined. Preliminary plastic strain at negligible reduction is shown to accelerate growth of the zone of internal nitridation and decelerates growth of the nitride zone. Nitrogen atom removal from the surface to the centre is retarded at the increase of the degree of reduction up to 50% and higher. The degree of deformations is the higher the lower nitridation temperature is

Processing of Refractory Metal Alloys for JOYO Irradiations

International Nuclear Information System (INIS)

RF Luther; ME Petrichek

2006-01-01

This is a summary of the refractory metal processing experienced by candidate Prometheus materiats as they were fabricated into specimens destined for testing within the JOYO test reactor, ex-reactor testing at Oak Ridge National Laboratory (ORNL), or testing within the NRPCT. The processing is described for each alloy from the point of inception to the point where processing was terminated due to the cancellation of Naval Reactor's involvement in the Prometheus Project. The alloys included three tantalum-base alloys (T-111, Ta-10W, and ASTAR-811C), a niobium-base alloy, (FS-85), and two molybdenum-rhenium alloys, one containing 44.5 w/o rhenium, and the other 47.5 w/o rhenium. Each of these alloys was either a primary candidate or back-up candidate for cladding and structural applications within the space reactor. Their production was intended to serve as a forerunner for large scale production ingots that were to be procured from commercial refractory metal vendors such as Wah Chang

Annex 4 - Task 08/13 final report, Producing the binary uranium alloys with alloying components Al, Mo, Zr, Nb, and B

International Nuclear Information System (INIS)

Lazarevic, Dj.

1961-01-01

Due to reactivity of uranium in contact with the gasses O 2 , N 2 , H 2 , especially under higher temperatures uranium processing is always done in vacuum or inert gas. Melting, alloying and casting is done in high vacuum stoves. This report reviews the type of furnaces and includes detailed description of the electric furnace for producing uranium alloys which is available in the Institute

The use of atomic absorption spectroscopy to measure arsenic, selenium, molybdenum, and vanadium in water and soil samples from uranium mill tailings sites

International Nuclear Information System (INIS)

Hollenbach, M.H.

1988-01-01

The Technical Measurements Center (TMC) was established to support the environmental measurement needs of the various DOE remedial action programs. A laboratory intercomparison study conducted by the TMC, using soil and water samples from sites contaminated by uranium mill tailings, indicated large discrepancies in analytical results reported by participating laboratories for arsenic, selenium, molybdenum, and vanadium. The present study was undertaken to investigate the most commonly used analytical techniques for measuring these four elements, ascertain routine and reliable quantification, and assess problems and successes of analysts. Based on a survey of the technical literature, the analytical technique of atomic absorption spectroscopy was selected for detailed study. The application of flame atomic absorption, graphite furnace atomic absorption, and hydride generation atomic absorption to the analysis of tailings-contaminated samples is discussed. Additionally, laboratory sample preparation methods for atomic absorption spectroscopy are presented. The conclusion of this report is that atomic absorption can be used effectively for the determination of arsenic, selenium, molybdenum, and vanadium in water and soil samples if the analyst understands the measurement process and is aware of potential problems. The problem of accurate quantification of arsenic, selenium, molybdenum, and vanadium in water and soil contaminated by waste products from uranium milling operations affects all DOE remedial action programs [Surplus Facilities Management Program (SFMP), Formerly Utilized Site Remedial Action Program (FUSRAP), and Uranium Mill Tailings Remedial Action Program (UMTRAP)], since all include sites where uranium was processed. 96 refs., 9 figs

Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

International Nuclear Information System (INIS)

Loelgen, R.H.; Cundy, M.R.; Schuele, W.

1977-01-01

A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this observation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. The diffusion tests will be performed at the Euratom Joint Research Centre in Ispra, Italy, and the irradiation creep tests will be carried out in the High Flux Reactor /9/ of the Euratom Joint Research Centre in Petten, The Netherlands. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. In the first type of rig about 50 samples can be tested uniaxially under tension with various combinations of irradiation temperature and stress. The second type of rig holds up to 70 samples which are tested in bending, again with various combinations of irradiation temperature and stress

The dangers of irradiate uranium in nuclear reactors; Les dangers de l'uranium irradie dans les reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Jammet, H; Joffre, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The danger of the uranium cans sur-activated by the use in the nuclear reactors is triple: - Irradiation from afar, during manipulations of the cans. - Contamination of air when decladding. - Contamination of air by fire of uranium in a reactor in function The first two dangers are usual and can be treated thanks to the rules of security in use in the atomic industry. The third has an accidental character and claimed for the use of special and exceptional rules, overflowing the industrial setting, to reach the surrounding populations. (authors) [French] Le danger des cartouches d'uranium suractive par utilisation dans les reacteurs nucleaires est triple: - Irradiation a distance, lors des manipulations des cartouches. - Contamination de l'air au moment de leur degainage. - Contamination de l'air par incendie d'uranium dans un reacteur en fonctionnement. Les deux premiers dangers sont habituels et peuvent etre traites grace aux regles de securite en usage dans l'industrie atomique. Le troisieme revet un caractere accidentel et reclame l'emploi de regles speciales et exceptionnelles, debordant le cadre industriel, pour atteindre celui des populations environnantes. (auteurs)

Some potential strategies for the treatment of waste uranium metal and uranium alloys

International Nuclear Information System (INIS)

Burns, C.J.; Frankcom, T.M.; Gordon, P.L.; Sauer, N.N.

1993-01-01

Large quantities of uranium metal chips and turnings stored throughout the DOE Complex represent a potential hazard, due to the reactivity of this material toward air and water. Methods are being sought to mitigate this by conversion of the metal, via room temperature solutions routes, to a more inert oxide form. In addition, the recycling of uranium and concomitant recovery of alloying metals is a desirable goal. The emphasis of the authors' research is to explore a variety of oxidation and reduction pathways for uranium and its compounds, and to investigate how these reactions might be applied to the treatment of bulk wastes

Nuclear criticality safety parameter evaluation for uranium metallic alloy

Energy Technology Data Exchange (ETDEWEB)

Sanchez, Andrea; Abe, Alfredo, E-mail: andreasdpz@hotmail.com, E-mail: abye@uol.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Energia Nuclear

2013-07-01

Nuclear criticality safety during fuel fabrication process, transport and storage of fissile and fissionable materials requires criticality safety analysis. Normally the analysis involves computer calculations and safety parameters determination. There are many different Criticality Safety Handbooks where such safety parameters for several different fissile mixtures are presented. The handbooks have been published to provide data and safety principles for the design, safety evaluation and licensing of operations, transport and storage of fissile and fissionable materials. The data often comprise not only critical values, but also subcritical limits and safe parameters obtained for specific conditions using criticality safety calculation codes such as SCALE system. Although many data are available for different fissile and fissionable materials, compounds, mixtures, different enrichment level, there are a lack of information regarding a uranium metal alloy, specifically UMo and UNbZr. Nowadays uranium metal alloy as fuel have been investigated under RERTR program as possible candidate to became a new fuel for research reactor due to high density. This work aim to evaluate a set of criticality safety parameters for uranium metal alloy using SCALE system and MCNP Monte Carlo code. (author)

Irradiation effects in magnesium and aluminium alloys

International Nuclear Information System (INIS)

Sturcken, E.F.

1979-01-01

Effects of neutron irradiation on microstructure, mechanical properties and swelling of several magnesium and aluminium alloys were studied. The neutron fluences of 2-3 X 10 22 n/cm 2 , >0.2 MeV produced displacement doses of 20 to 45 displacements per atom (dpa). Ductility of the magnesium alloys was severely reduced by irradiation induced recrystallization and precipitation of various forms. Precipitation of transmuted silicon occurred in the aluminium alloys. However, the effect on ductility was much less than for the magnesium alloys. The magnesium and aluminium alloys had excellent resistance to swelling: The best magnesium alloy was Mg/3.0 wt% Al/0.19 wt% Ca; its density decreased by only 0.13%. The best aluminium alloy was 6063, with a density decrease of 0.22%. (Auth.)

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

Energy Technology Data Exchange (ETDEWEB)

Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; 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. - Highlights: •Complementary fission gas release events are reported for U-Mo fuel with and without cladding. •Exothermic reaction between Zr diffusion layer and cladding influences fission gas release. •Mechanisms responsible for fission gas release are similar, but with varying timing and magnitude. •Behavior of samples is similar after 800 °C signaling the onset of superlattice destabilization.

A Very High Uranium Density Fission Mo Target Suitable for LEU Using atomization Technology

Energy Technology Data Exchange (ETDEWEB)

Kim, C. K.; Kim, K. H.; Lee, Y. S.; Ryu, H. J.; Woo, Y. M.; Jang, S. J.; Park, J. M.; Choi, S. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

Currently HEU minimization efforts in fission Mo production are underway in connection with the global threat reduction policy. In order to convert HEU to LEU for the fission Mo target, higher uranium density material could be applied. The uranium aluminide targets used world widely for commercial {sup 99}Mo production are limited to 3.0 g-U/cc in uranium density of the target meat. A consideration of high uranium density using the uranium metal particles dispersion plate target is taken into account. The irradiation burnup of the fission Mo target are as low as 8 at.% and the irradiation period is shorter than 7 days. Pure uranium material has higher thermal conductivity than uranium compounds or alloys. It is considered that the degradation by irradiation would be almost negligible. In this study, using the computer code of the PLATE developed by ANL the irradiation behavior was estimated. Some considerations were taken into account to improve the irradiation performance further. It has been known that some alloying elements of Si, Cr, Fe, and Mo are beneficial for reducing the swelling by grain refinement. In the RERTR program recently the interaction problem could be solved by adding a small amount of Si to the aluminum matrix phase. The fabrication process and the separation process for the proposed atomized uranium particles dispersion target were reviewed

Surface preparation process of a uranium titanium alloy, in particular for chemical nickel plating

International Nuclear Information System (INIS)

Henri, A.; Lefevre, D.; Massicot, P.

1987-01-01

In this process the uranium alloy surface is attacked with a solution of lithium chloride and hydrochloric acid. Dissolved uranium can be recovered from the solution by an ion exchange resin. Treated alloy can be nickel plated by a chemical process [fr

Microstructural investigation of as-cast uranium rich U–Zr alloys

Energy Technology Data Exchange (ETDEWEB)

Zhang, Yuting, E-mail: zhangyuting@caep.cn [Science and Technology on Surface Physics and Chemistry Laboratory, Jiangyou 621908, Sichuan (China); School of Nuclear Science and Technology, National Synchrotron Radiation Laboratory, University of Science and Technology of China, Hefei 230029, Anhui (China); Wang, Xin [Science and Technology on Surface Physics and Chemistry Laboratory, Jiangyou 621908, Sichuan (China); Zeng, Gang [Institute of Materials, China Academy of Engineering Physics, Jiangyou 621908, Sichuan (China); Wang, Hui [Science and Technology on Surface Physics and Chemistry Laboratory, Jiangyou 621908, Sichuan (China); Jia, Jianping [Institute of Materials, China Academy of Engineering Physics, Jiangyou 621908, Sichuan (China); Sheng, Liusi [School of Nuclear Science and Technology, National Synchrotron Radiation Laboratory, University of Science and Technology of China, Hefei 230029, Anhui (China); Zhang, Pengcheng, E-mail: zpc113@sohu.com [Science and Technology on Surface Physics and Chemistry Laboratory, Jiangyou 621908, Sichuan (China)

2016-04-01

The present study evaluates the microstructure in as-cast uranium rich U–Zr alloys, an important subsystem of U–Pu–Zr ternary metallic nuclear reactor fuel, as a function of the Zr content, from 2wt.% to 15wt.%Zr. It has been previously suggested that the unique intermetallic compound δ phase in U–Zr alloys is only present in as-cast U–Zr alloys with a Zr content exceeding 10wt.%Zr. However, our analysis of transmission electron microscopy (TEM) data shows that the δ phase is common to all as-cast alloys studied in this work. Furthermore, specific coherent orientation relationship is found between the α and δ phases, consistent with previous findings, and a third variant is discovered in this paper. - Highlights: • Initially, lattice parameter of as-cast U–Zr alloys decrease with the increasing Zr content, and then increase. • XRD data show the presence of δ-UZr{sub 2} phase in as-cast U–Zr alloys with a Zr content of more than 8wt.% Zr. • Finding δ-UZr{sub 2} phase exists in all as-cast uranium rich U–Zr alloys, even for alloys with a lean Zr content. • Three kinds of preferential orientations of the δ phase grow.

Influence of irradiation and radiolysis on the corrosion rates and mechanisms of zirconium alloys

International Nuclear Information System (INIS)

Verlet, Romain

2015-01-01

The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO 2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO 2 pellets + cladding). To optimize core management and to extend the lifetime of the fuel rods in reactor, new alloys based on zirconium-niobium (M5) have been developed. However, the corrosion mechanisms of these are not completely understood because of the complexity of these materials, corrosion environment and the presence of radiation from the nuclear fuel. Therefore, this thesis specifically addresses the effects of radiolysis and defects induced by irradiation with ions in the matrix metal and the oxide layer on the corrosion rate of Zircaloy-4 and M5. The goal is to separate the influence of radiation damage to the metal, that relating to defects created in the oxide and that linked to radiolysis of the primary medium on the oxidation rate of zirconium alloys in reactor. 1) Regarding effect of irradiation of the metal on the oxidation rate: type dislocation loops appear and increase the oxidation rate of the two alloys. For M5, in addition to the first effect, a precipitation of fines needles of niobium reduced the solid solution of niobium concentration in the metal and ultimately in the oxide, which strongly reduces the oxidation rate of the alloy. 2) Regarding the effect of irradiation of the oxide layer on the oxidation rate: defects generated by the nuclear cascades in the oxide increase the oxidation rate of the two materials. For M5, germination of niobium enriched zones in irradiated oxide also causes a decrease of the niobium concentration in solid solution

Irradiation of aluminium alloy materials with electron beam

International Nuclear Information System (INIS)

Konno, Osamu; Masumoto, Kazuyoshi

1982-01-01

It is a theme with a room for discussion to employ the stainless steel composed of longer half-life materials for the vacuum system of accelerators, from the viewpoint of radiation exposure. Therefore, it is desirable to use aluminium of shorter half-life in place of stainless steel. As a result of investigation on the above theme in the 1.2 GeV electron linac project in Tohoku University, it has been concluded that aluminium alloy vacuum chambers can reduce exposure dose by about one or two figures as compared with stainless steel ones. Of course, aluminium alloy contains trace amounts of Mg, Si, Ti, Cr, Mn, Fe, Zn, Cu and others. Therefore, four kinds of aluminium alloy considered to be usable have been examined for induced radioactivity by electron beam irradiation. Stainless steel SUS 304 has been also irradiated for comparison. Radiation energy has been 30 MeV and 200 MeV. When stainless steel and aluminium alloy were compared, aluminium alloy was very effective for reducing surface dose in low energy irradiation. In 200 MeV irradiation, the dose ratio of aluminium alloy to stainless steel became 1/30 to 1/100 after one week, though the dose difference between these two materials became smaller in 100 days or more after irradiation. If practical inspection and repair are implemented during the period from a few days to one week after shutdown, the aluminium alloy is preferable for exposure dose reduction even in high energy irradiation. (Wakatsuki, Y.)

Corrosion characteristics of Hastelloy N alloy after He+ ion irradiation

International Nuclear Information System (INIS)

Lin Jianbo; Yu Xiaohan; Li Aiguo; He Shangming; Cao Xingzhong; Wang Baoyi; Li Zhuoxin

2014-01-01

With the goal of understanding the invalidation problem of irradiated Hastelloy N alloy under the condition of intense irradiation and severe corrosion, the corrosion behavior of the alloy after He + ion irradiation was investigated in molten fluoride salt at 700 °C for 500 h. The virgin samples were irradiated by 4.5 MeV He + ions at room temperature. First, the virgin and irradiated samples were studied using positron annihilation lifetime spectroscopy (PALS) to analyze the influence of irradiation dose on the vacancies. The PALS results showed that He + ion irradiation changed the size and concentration of the vacancies which seriously affected the corrosion resistance of the alloy. Second, the corroded samples were analyzed using synchrotron radiation micro-focused X-ray fluorescence, which indicated that the corrosion was mainly due to the dealloying of alloying element Cr in the matrix. Results from weight-loss measurement showed that the corrosion generally correlated with the irradiation dose of the alloy. (author)

Determination of crystalline texture in aluminium - uranium alloys by neutron diffraction

International Nuclear Information System (INIS)

Azevedo, A.M.V. de.

1978-01-01

Textures of hot-rolled aluminum-uranium alloys and of aluminum were determined by neutron diffraction. Sheets of alloys containing 8.0, 21.5 and 23.7 wt pct U, as well as pure aluminum, were obtained in a stepped rolling process, 15% reduction each step, 75% total reduction. During the rolling the temperature was 600 0 C. Alloys with low uranium contents are two phase systems in which an intermetallic compound UAl 4 , orthorhombic, is dispersed in a pure aluminum matrix. The addition of a few percent of Si in such alloys leads to the formation of UAl 3 , simple cubic, instead of UAl 4 . The Al -- 23.7 wt pct U alloy was prepared with 2,2 wt pct of Si. The results indicate that the texture of the matrix is more dependent on the uranium concentration than on the texture of the intermetallic phases. An improvement in the technique applied to texture measurements by using a sample fully bathed in the neutron beam is also presented. The method takes advantage of the low neutron absorption of the studied materials as well as of the neglibible variation in the multiple scattering which occurs in a conveniently shaped sample having a weakly developed texture. (Author) [pt

Analytical electron microscopy of neutron-irradiated reactor alloys

International Nuclear Information System (INIS)

Thomas, L.E.

1982-01-01

Exposure to the high neutron fluxes and temperatures from 400 to 650 0 C in the core region of a fast breeder reactor profoundly alters the microstructure and properties of structural steels and superalloys. The development of irradiation-induced voids, dislocations and precipitates, as well as segregation of alloying elements on a microscopic scale has been related to macroscopic swelling, creep, hardening and embrittlement which occur during prolonged exposures in reactor. Microanalytical studies using TEM/STEM methods, primarily energy dispersive x-ray (EDX) microanalysis, have greatly aided understanding of alloy behavior under irradiation. The main uses of analytical electron microscopy in studying irradiated alloys have been the identification of irradiation-induced precipitates and determination of the changes in local composition due to irradiation-induced solute segregation

RERTR-12 Insertion 1 Irradiation Summary Report

International Nuclear Information System (INIS)

Perez, D.M.; Lillo, M.A.; Chang, G.S.; Woolstenhulme, N.E.; Roth, G.A.; Wachs, D.M.

2012-01-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications. RERTR-12 insertion 1 includes the capsules irradiated during the first two irradiation cycles. These capsules include Z, X1, X2 and X3 capsules. The following report summarizes the life of the RERTR-12 insertion 1 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

RERTR-12 Insertion 2 Irradiation Summary Report

International Nuclear Information System (INIS)

Perez, D.M.; Chang, G.S.; Wachs, D.M.; Roth, G.A.; Woolstenhulme, N.E.

2012-01-01

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

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

International Nuclear Information System (INIS)

McDeavitt, Sean M.

2011-01-01

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

Contribution to the micrographic study of uranium and its alloys; Contribution a l'etude micrographique de l'uranium et de ses alliages

Energy Technology Data Exchange (ETDEWEB)

Monti, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1956-06-15

The present report is the result of research carried out by the radio metallurgy section, to perfect micrographic techniques applicable to the study of samples of irradiated uranium. In the first part of this work, two polishing baths are developed, having the qualities with a minimum of disadvantages inherent in their respective compositions: they are, on the one hand perchloric acid-ethanol mixtures, and on the other hand a phospho-chromic-ethanol bath. In the chapter following, the micrographic attack of uranium is studied. The only satisfactory process is oxidation by cathode bombardment forming epitaxic layers. In the third chapter, an attempt is made to characterise the different surface states of the uranium by dissolution potential measurements and electronic diffraction. In the fourth chapter are given some examples of the application of these techniques to the micrographic study of various uranium alloys. In an appendix, it is shown how the chemical oxidation after phospho-chromic-alcohol polishing allows the different inclusions present in the molten uranium to be distinguished. By X-ray diffraction, uranium monocarbide and mononitride inclusions in particular are characterised. (author) [French] Le present rapport est le resultat de recherches effectuees au service de radiometallurgie pour la mise au point de techniques micrographiques applicables a l'etude d'echantillons d'uranium irradie. Dans la premiere partie de ce travail, nous mettons au point deux bains de polissage qui presentent les qualites inherentes a leur composition respective, avec le minimum d'inconvenients: ce sont d'une part des melanges acide perchlorique-ethanol, et d'autre part un bain phospho-chromique-ethanol. Dans le chapitre suivant, nous etudions l'attaque micrographique de l'uranium. Seul le procede d'oxydation par bombardement cathodique formant des couches epitaxiques, est satisfaisant. Dans le troisieme chapitre, nous essayons de caracteriser les differents etats de

Postirradiation examination of high-density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Strain, R.V.

1998-01-01

Two irradiation test vehicles, designated RERTR-2, were inserted into the Advanced Test reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn: the intermetallic compounds U 2 Mo and U-10Mo-0.-5Sn; the intermetallic compounds U 2 Mo and U 3 Si 2 were also included in the fuel test matrix. These fuels are included in the experiments as microplates (76 mm x 22 mm x 1.3mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature (∼100 deg C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively at calculated peak fuel burnups of 45 and 71 at %-U 235 Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments. (author)

Feasibility of Producing Molybdenum-99 on a Small Scale Using Fission of Low Enriched Uranium or Neutron Activation of Natural Molybdenum

International Nuclear Information System (INIS)

2015-01-01

This publication documents the work performed within the IAEA coordinated research project (CRP) on Developing Techniques for Small Scale Indigenous Molybdenum-99 Production Using LEU Fission or Neutron Activation. The project allowed participating institutions to receive training and information on aspects necessary for starting production of molybdenum-99 ( 99 Mo) on a small scale, that is, to become national level producers of this medical isotope. Stable production of 99Mo is one of the most pressing issues facing the nuclear community at present, because the medical isotope technetium-99m ( 99m Tc), which decays from 99 Mo, is one of the most widely used radionuclides in diagnostic imaging and treatment around the world. In the past five years, there have been widespread shortages of 99 Mo owing to the limited number of producers, many of which use ageing facilities. To assist in stabilizing the production of 99Mo, and to promote the use of production methods that do not rely on the use of highly enriched uranium (HEU), the IAEA initiated the abovementioned CRP on small scale 99Mo production using low enriched uranium (LEU) fission or neutron activation methods. The intention was to enable participating institutions to gain the knowledge necessary to become national level producers of 99Mo in the event of further global shortages. Some of the institutions that participated in the CRP have continued their work on 99 Mo production, and are enlisting the assistance of other CRP members and the IAEA’s technical cooperation programme to set up a small scale production capability. In total, the CRP was active for six years, and concluded in December 2011. During the CRP, fourteen IAEA Member States took part; four research coordination meetings were held, and four workshops were held on operational aspects of 99 Mo production, LEU target fabrication and waste management. Most participants carried out work related to the entire production process, from target

Atmospheric corrosion of uranium-carbon alloys

International Nuclear Information System (INIS)

Rousset, P.; Accary, A.

1965-01-01

The authors study the corrosion of uranium-carbon alloys having compositions close to that of the mono-carbide; they show that the extent of the observed corrosion effects increases with the water vapour content of the surrounding gas and they conclude that the atmospheric corrosion of these alloys is due essentially to the humidity of the air, the effect of the oxygen being very slight at room temperature. They show that the optimum conditions for preserving U-C alloys are either a vacuum or a perfectly dry argon atmosphere. The authors have also established that the type of corrosion involved is a corrosion which 'cracks under stress' and is transgranular (it can also be intergranular in the case of sub-stoichiometric alloys). They propose, finally, two hypotheses for explaining this mechanism, one of which is illustrated by the existence, at the fissure interface, of corrosion products which can play the role of 'corners' in the mono-carbide grains. (authors) [fr

Neutron resistant irradiation alloy and usage thereof

International Nuclear Information System (INIS)

Okada, Osamu; Nakata, Kiyotomo; Kato, Takahiko.

1997-01-01

A neutron irradiation embrittlement-resistant alloy comprising a Ti alloy having an average grain size of 2μm or smaller and containing from 30 to 40wt% of Al is subjected to powder solidification and then to isothermal forging at a forging rate of from 50 to 80% at a temperature range of from 1150 to 1500K. Namely, since the Ti-Al type alloy comprises from 30 to 30wt% of Al, optionally, from 1 to 6% of Mn, from 0.1 to 0.5% of Si, from 4 to 16% of V and the balance of Ti, it has excellent specific strength, high durable temperature and excellent neutron irradiation resistance, and has ductility required as structural materials. Accordingly, if the Ti-Al type alloy excellent in embrittlement resistance to neutron irradiation dimensional stability of materials is applied to constitutional parts of a reactor core of a nuclear reactor and a thermonuclear reactor to be exposed under neutron irradiation, high reliability is provided and the amount of activated materials is reduced by improving the working life of the materials. (N.H.)

Fusion neutron irradiation of Ni(Si) alloys at high temperature

International Nuclear Information System (INIS)

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

1987-09-01

Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab

Fusion neutron irradiation of Ni(Si) alloys at high temperature

Energy Technology Data Exchange (ETDEWEB)

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

1987-09-01

Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab.

Recovery of electron irradiated V-Ga alloys

International Nuclear Information System (INIS)

Leguey, T.; Monge, M.; Pareja, R.; Hodgson, E.R.

2000-01-01

The recovery characteristics of electron-irradiated V-Ga alloys with 1.2 and 4.6 at.% Ga have been investigated by positron annihilation spectroscopy (PAS). It is found that vacancies created by electron irradiation become mobile in these alloys at ∼293 K. This temperature is noticeably lower than that in pure V and V-Ti alloys. The vacancies aggregate into microvoids in V-4.6Ga, but do not in V-1.2Ga. The results indicate that vacancies are bound to Ga-interstitial impurity pairs

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

Energy Technology Data Exchange (ETDEWEB)

McDeavitt, Sean M

2011-04-29

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

Irradiation behavior of miniature experimental uranium silicide fuel plates

International Nuclear Information System (INIS)

Hofman, G.L.; Neimark, L.A.; Mattas, R.F.

1983-01-01

Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 10 20 cm -3 , far short of the approximately 20 x 10 20 cm -3 goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix

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

Science.gov (United States)

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

2018-04-01

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

Development of very high-density low-enriched uranium fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

1997-02-01

The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

Thermal analysis of LEU modified Cintichem target irradiated in TRIGA reactor

International Nuclear Information System (INIS)

Catana, A; Toma, C.

2009-01-01

Actions conceived during last years at international level for conversion of Molybdenum fabrication process from HEU to LEU targets utilization created opportunities for INR to get access to information and participating to international discussions under IAEA auspices. Concrete steps for developing fission Molybdenum technology were facilitated. Institute of Nuclear Research bringing together a number of conditions like suitable irradiation possibilities, direct communication between reactor and hot cell facility, handling capacity of high radioactive sources, and simultaneously the existence of an expanding internal market, decided to undertake the necessary steps in order to produce fission molybdenum. Over the course of last years of efforts in this direction we developed the steps for fission Molybdenum technology development based on modified Cintichem process in accordance with the Argonne National Laboratory proved methodology. Progress made by INR to heat transfer computations of annular target using is presented. An advanced thermal-hydraulic analysis was performed to estimate the heat removal capability for an enriched uranium (LEU) foil annular target irradiated in TRIGA reactor core. As a result, the present analysis provides an upper limit estimate of the LEU-foil and external target surface temperatures during irradiation in TRIGA 14 MW reactor. (authors)

Proton irradiation studies on Al and Al5083 alloy

Science.gov (United States)

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

2017-10-01

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

Neutron irradiation effect on thermomechanical properties of shape memory alloys

International Nuclear Information System (INIS)

Abramov, V.Ya.; Ionajtis, R.R.; Kotov, V.V.; Loguntsev, E.N.; Ushakov, V.P.

1996-01-01

Alloys of Ti-Ni, Ti-Ni-Pd, Fe-Mn-Si, Mn-Cu-Cr, Mn-Cu, Cu-Al-Mn, Cu-Al-Ni systems are investigated after irradiation in IVV-2M reactor at various temperatures with neutron fluence of 10 19 - 10 20 cm -2 . The degradation of shape memory effect in titanium nickelide base alloys is revealed after irradiation. Mn-Cu and Mn-Cu-Cr alloys show the best results. Trends in shape memory alloy behaviour depending on irradiation temperature are found. A consideration is given to the possibility of using these alloys for components of power reactor control and protection systems [ru

Development of metal uranium fuel and testing of construction materials (I-VI); Part I; Razvoj metalnog goriva i ispitivanje konstrukcionih materijala (I-VI deo); I deo

Energy Technology Data Exchange (ETDEWEB)

Mihajlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-11-15

This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors.

Contribution to the study of the fission-gas release in metallic nuclear fuels

International Nuclear Information System (INIS)

Kryger, B.

1969-10-01

In order to study the effect of an external pressure on the limitation of swelling due to fission-gas precipitation, some irradiations have been carried out at burn-ups of about 35.000 MWd/ton, and at average sample temperatures of 575 Celsius degrees, of non-alloyed uranium and uranium 8 per cent molybdenum gained in a thick stainless steel can. A cylindrical central hole allows a fuel swelling from 20 to 33 per cent according to the experiment. After irradiation, the uranium samples showed two types of can rupture: one is due to the fuel swelling, and the other, to the pressure of the fission gases, released through a network of microcracks. The cans of the uranium-molybdenum samples are all undamaged and it is shown that the gas release occurs by interconnection of the bubbles for swelling values higher than those obtained in the case of uranium. For each type of fuel, a swelling-fission gas release relationship is established. The results suggest that good performances with a metallic fuel intended for use in fast reactor conditions can be obtained. (author) [fr

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

Science.gov (United States)

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

1988-07-01

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

Experiments for separation and purification of Mo-99 from uranium solution with fission products as tracers

International Nuclear Information System (INIS)

Androne, Gabriela Elena; Matei, Ana

2007-01-01

Technetium-99 m ( 99m Tc, T 1/2 = 6 hours), one of the most utilised radioisotopes in nuclear medicine, is generated through the beta decay of 99 Mo (T 1/2 = 66 hours) and which will decay through isomer transition to 99 Tc ( T 1/2 = 2.10 5 years) through the emission of a gamma radiation with the energy of 0.140 MeV. The work presents the phases of the process of Mo separation and purification at a tracer level. The tests performed in the laboratory have established the optimum conditions for the separation and purification of Molybdenum. To establish the separation and purification parameters, a synthetic solution which contains the elements which result following the irradiation of a low enriched Uranium foil weighing 10 g (∼20 % 235 U). To mark this solution, about 13 mg of UO 2 10% 235 U was irradiated for 2000 s, at a flux of about 7x10 12 n/cm 2 s. This amount of UO 2 will be added to the above-mentioned solution after dissolution. The method for separating the Molybdenum from irradiated Uranium solution is one of selective precipitation of Mo with α-benzoin-oxyme (α- BO). To purify the Molybdenum solution, two purification columns were utilised. Their role was to absorb the impurities remained in the mass of the precipitate. They and the Molybdenum have passed into the solution simultaneously, allowing the Molybdenum to pass. These columns are: the column with active charcoal (AC)+ active charcoal covered with silver (AgAC); active charcoal column (AC)+active charcoal covered with silver (AgAC)+hydrated zirconium oxide (HZO). Although all the phases of the process are performed with high yields, the final yields of recovery of Mo from U solutions are higher than 80%. (authors)

Metallurgical processing of the uranium-0.75 titanium alloy

International Nuclear Information System (INIS)

Jessen, N.C.

1976-01-01

Although the addition of titanium is an effective means of strengthening uranium, careful control of casting, homogenization, and heat treatment are necessary to optimize mechanical properties. Quenching of the alloy provides increased strength and elongation; however, subsequent low temperature aging will increase the strength even higher at the sacrifice of ductility. The properties of the alloy are quench rate sensitive and quenching produces high residual stresses in the alloy. The residual stresses can be reduced by mechanical deformation with only slight degradation of the mechanical properties. 15 figures

Analysis of uranium and of some of its compounds and alloys. Copper spectrophotometric determination

International Nuclear Information System (INIS)

Copper determination in uranium, uranium oxides (UO 2 , UO 3 , U 3 O 8 ), ammonium diuranate, U-Al-Fe alloy (700 ppm Al and 300 ppm Fe) and U-Mo alloy (1.1 percent Mo) by acid dissolution reduction of copper by hydroxylamine hydrochloride and formation of a complex with diquinolyle-2,2' amyl alcohol (pH value 6 to 7) and spectrophotometry at 550 nm. The method is applicable for copper content between 5 to 40 ppm in respect of uranium contained in the material [fr

Contribution to the micrographic study of uranium and its alloys; Contribution a l'etude micrographique de l'uranium et de ses alliages

Energy Technology Data Exchange (ETDEWEB)

Monti, H. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1956-06-15

The present report is the result of research carried out by the radio metallurgy section, to perfect micrographic techniques applicable to the study of samples of irradiated uranium. In the first part of this work, two polishing baths are developed, having the qualities with a minimum of disadvantages inherent in their respective compositions: they are, on the one hand perchloric acid-ethanol mixtures, and on the other hand a phospho-chromic-ethanol bath. In the chapter following, the micrographic attack of uranium is studied. The only satisfactory process is oxidation by cathode bombardment forming epitaxic layers. In the third chapter, an attempt is made to characterise the different surface states of the uranium by dissolution potential measurements and electronic diffraction. In the fourth chapter are given some examples of the application of these techniques to the micrographic study of various uranium alloys. In an appendix, it is shown how the chemical oxidation after phospho-chromic-alcohol polishing allows the different inclusions present in the molten uranium to be distinguished. By X-ray diffraction, uranium monocarbide and mononitride inclusions in particular are characterised. (author) [French] Le present rapport est le resultat de recherches effectuees au service de radiometallurgie pour la mise au point de techniques micrographiques applicables a l'etude d'echantillons d'uranium irradie. Dans la premiere partie de ce travail, nous mettons au point deux bains de polissage qui presentent les qualites inherentes a leur composition respective, avec le minimum d'inconvenients: ce sont d'une part des melanges acide perchlorique-ethanol, et d'autre part un bain phospho-chromique-ethanol. Dans le chapitre suivant, nous etudions l'attaque micrographique de l'uranium. Seul le procede d'oxydation par bombardement cathodique formant des couches epitaxiques, est satisfaisant. Dans le troisieme chapitre, nous essayons

Low temperature irradiation effects on iron boron based amorphous metallic alloys

International Nuclear Information System (INIS)

Audouard, A.

1982-09-01

Three Fe-B amorphous alloys (Fe 80 B 20 , Fe 27 Mo 2 B 20 and Fe 75 B 25 ) and the crystallized Fe 3 B alloy have been irradiated at the temperature of liquid hydrogen. Electron irradiation and irradiation by 10 B fission fragments induce point defects in amorphous alloys. These defects are characterized by an intrinsic resistivity and a formation volume. The threshold energy for the displacement of iron atoms has also been calculated. Irradiation by 235 U fission fragments induces some important structural modifications in the amorphous alloys [fr

Irradiation of copper alloys in FFTF

International Nuclear Information System (INIS)

Brager, H.R.; Garner, F.A.

1984-01-01

Nine copper-base alloys in thirteen material conditions have been inserted into the MOTA-18 experiment for irradiation in FFTF at approx.450 0 C. The alloy Ni-1.9Be is also included in this experiment, which includes both TEM disks and miniature tensile specimens

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

Science.gov (United States)

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.

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nikel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) seamless pipe and tube

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nikel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) seamless pipe and tube

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045 and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) plate, sheet and strip

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045 and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) plate, sheet and strip

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) rod, bar, and wire

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) rod, bar, and wire

Mechanical and wear properties of pre-alloyed molybdenum P/M steels with nickel addition

Directory of Open Access Journals (Sweden)

Yamanoglu R.

2012-01-01

Full Text Available The aim of this study is to understand the effect of nickel addition on mechanical and wear properties of molybdenum and copper alloyed P/M steel. Specimens with three different nickel contents were pressed under 400 MPa and sintered at 1120ºC for 30 minutes then rapidly cooled. Microstructures and mechanical properties (bending strength, hardness and wear properties of the sintered specimens were investigated in detail. Metallographical investigations showed that the microstructures of consolidated specimens consist of tempered martensite, bainite, retained austenite and pores. It is also reported that the amount of pores varies depending on the nickel concentration of the alloys. Hardness of the alloys increases with increasing nickel content. Specimens containing 2% nickel showed minimum pore quantity and maximum wear resistance. The wear mechanism changed from abrasive wear at low nickel content to adhesive wear at higher nickel content.

Irradiation of steel, molybdenum and tungsten - VISA-2f; Ozracivanje celika, molibdena i volframa - VISA-2f -

Energy Technology Data Exchange (ETDEWEB)

Veljkovic, S; Milasin, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1962-12-15

The objective of the experiment is to study the radiation damage of steel, molybdenum and tungsten after irradiation under fast neutron flux. The sample wires of steel Mo and W will be irradiated, integral fast neutron flux should be higher than 10{sup 18} neutrons/cm{sup 2}, the temperature should be as low as possible.

Development of very-high-density low-enriched uranium fuels

International Nuclear Information System (INIS)

Snegrove, J.L.; Hofmann, G.L.; Trybus, C.L.; Wiencek, T.C.

1997-01-01

The RERTR (=Reduced Enrichment for Research and Test Reactors) program has begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun. (author)

Powder production of U-Mo alloy, HMD process (Hydriding- Milling- Dehydriding)

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, E. E.; Garcia, J.H.; Lopez, M.; Cabanillas, E.; Adelfang, P. [Dept. Combustibles Nucleares. Comision Nacional de Energia Atomica, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina)

2002-07-01

Uranium-molybdenum (U-Mo) alloys can be hydrided massively in metastable {gamma} (gamma) phase. The brittle hydride can be milled and dehydrided to acquire the desired size distributions needed for dispersion nuclear fuels. The developments of the different steps of this process called hydriding-milling- dehydriding (HMD Process) are described. Powder production scales for industrial fabrication is easily achieved with conventional equipment, small man-power and low investment. (author)

Powder production of U-Mo alloy, HMD process (Hydriding- Milling- Dehydriding)

International Nuclear Information System (INIS)

Pasqualini, E. E.; Garcia, J.H.; Lopez, M.; Cabanillas, E.; Adelfang, P.

2002-01-01

Uranium-molybdenum (U-Mo) alloys can be hydrided massively in metastable γ (gamma) phase. The brittle hydride can be milled and dehydrided to acquire the desired size distributions needed for dispersion nuclear fuels. The developments of the different steps of this process called hydriding-milling- dehydriding (HMD Process) are described. Powder production scales for industrial fabrication is easily achieved with conventional equipment, small man-power and low investment. (author)

Preliminary study of the preparation of uranium 232 by irradiation of protactinium 231; Etude preliminaire a la preparation d'uranium 232 par irradiation de protactinium 231

Energy Technology Data Exchange (ETDEWEB)

Guillot, Ph. [Commissariat a l' Energie Atomique, Fontenay aux Roses (France). Centre d' Etudes Nucleaires

1965-07-01

A bibliography about preparation of uranium 232 is done. This even-even isotope of uranium is suitable for radioactive tracer, neutron source through {alpha},n reaction and heat source applications. The irradiation of protactinium 231, the chemical separation and the purification of uranium are studied. (author) [French] Une etude bibliographique de la preparation d'uranium 232 a ete effectuee. Cet isotope pair-pair de l'uranium peut etre utilise en tant que traceur, source d'energie et source de neutrons, lorsqu'il est melange a un element leger tel le beryllium. Une etude du taux de formation des isotopes produits, lors de l'irradiation du protactinium 231 - une des manieres d'obtenir l'uranium 232 - a ete faite a l'aide d'un programme passe sur ordinateur. Les problemes poses par la separation chimique et la purification de l'uranium ont ete egalement envisages dans ce rapport. (auteur)

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

Energy Technology Data Exchange (ETDEWEB)

Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Company, Richland, WA (United States)

1997-04-01

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

Uranium alloys for using in fast breeder reactors

International Nuclear Information System (INIS)

Moura Neto, C.; Pires, O.S.

1988-08-01

The U-Zr and U-Ti alloys are studied, given emphasis to the high solute solubility in gamma phase of uranium, which is suitable for using as metal fuel in fast breeder reactors. The alloys were prepared in electron beam furnaces and submitted to X-ray diffraction, X-ray fluorescence, microhardness, optical metallography, and chemical analysis. The obtained values are good agreements with the literature data. The study shows that the U-Zr presents better characteristics than the U-Ti for using as fuel in fast breeder reactors. (M.C.K.) [pt

Experimental measurement of fission fragments paths in uranium gold, molybdenum, zirconium and silicon; Mesure experimentale des parcours des fragments de fission dans l'uranium, l'or, le molybdene, le zirconium et le silicium

Energy Technology Data Exchange (ETDEWEB)

Faraggi, H; Garin-Bonnet, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The measurement of total number of fissiongments emerging from an homogeneous, thick alloy composed of uranium plus another element (the concentration of uranium being known) allows to obtain the range of the fragments in this alloy. By varying the concentration, the range of the fragments in uranium and in the other element can be deduced. (author)Fren. [French] La mesure du nombre total de fragments de fission sortant d'un alliage homogene epais d'uranium et d'un autre element, pour lequel la concentration en uranium est donnee, permet la mesure du parcours des fragments dans cet alliage. En faisant varier la concentration, on peut deduire de ces mesures le parcours des fragments dans l'uranium et dans l'autre element. (auteur)

Effects of exposure to high-temperature helium containing oxygen on the mechanical properties of molybdenum and TZM-Mo alloy at room temperature

International Nuclear Information System (INIS)

Noda, T.; Okada, M.; Watanabe, R.

1980-01-01

The effects of exposure to helium containing oxygen of 0.1-115 vpm at 1000 0 C on the mechanical properties of molybdenum and TZM-Mo alloy at room temperature were studied. The stress-relieved molybdenum specimen which was not recrystallized at test temperature showed the ductility after exposure to helium containing oxygen. The recrystallized molybdenum and TZM lost ductility after exposure to helium containing oxygen of 0.1-13 vpm in a few hours. The embrittlement of molybdenum was considered to be due to the grain boundary weakening. Molybdenum to which carbon was added seemed to hinder the grain boundary weakening by the oxygen contamination. Both stress-relieved and recrystallized TZM specimens picked up oxygen linearly with time of exposure to helium. The increase in oxygen content of TZM, which was considered to be caused by the internal oxidation of titanium and zirconium, results in the embrittlement of TZM. (orig.)

Progress in developing very-high-density low-enriched-uranium fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Hayes, S.L.; Wiencek, T.C.; Strain, R.V.

1999-01-01

Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm 3 . Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 deg. C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 deg. C in 8-g U/cm 3 fuel. (author)

Development of Molybdenum Adsorbent for 99Mo/99mTc Radioisotope Generator Based on Irradiated Natural Molybdenum

International Nuclear Information System (INIS)

Rohadi Awaludin; Hotman Lubis; Sriyono; Abidin; Herlina; Endang Sarmini; Indra Saptiama; Hambali

2011-01-01

Preparation of 99 Mo/ 99m Tc radioisotope generator using irradiated natural molybdenum requires an adsorbent with high absorption capacity. Zirconium-based materials (ZBM), adsorbent with adsorption capacity of about 183 mg(Mo) / g(adsorbent), has been successfully synthesized. However, the adsorbent was easily broken in the Mo adsorption process due to many fractures in the grain. To increase the hardness, the material was immersed in tetraethyl orthosilicate (TEOS) and coated by TEOS flow in a column. The hardness test results showed that the ZBM with TEOS treatment was not broken when immersed into the Mo solution. Observations using SEM showed that the fractures formed on the ZBM were successfully removed by TEOS treatment. Measurements using EDS showed that after TEOS treatment, the silicon was detected and the oxygen content increased in the material surface. Adsorption test results showed that the TEOS immersion decreased the adsorption capacity of molybdenum from 183 to 79.8 mg of Mo per gram of adsorbent. The TEOS flow-in a column gave material with relatively high adsorption capacity, 140 mgMo per gram adsorbent. The content of Silicon in the surface was lower than that of adsorbent immersed in TEOS. (author)

Uranium doping and neutron irradiation of Bi-2223 superconduction tapes for improved critical current density

International Nuclear Information System (INIS)

Moss, S.D.; Wang, W.G.; Dou, S.X.; Weinstein, R.

1998-01-01

It is demonstrated that a combination of neutron irradiation with uranium doping introduce fission tracks through a Bi-2223 tape which act as effective pinning centres, leading to a substantial increase in critical current. Preliminary data suggests that the combination of uranium doping and neutron irradiation produces improved flux pinning in Bi-2223 tapes over neutron irradiation alone. Before irradiation, SEM, DTA and XRD analyses were performed on the tapes. Both before and after irradiation the trapped maximum magnetic flux was measured at 77K. Before neutron irradiation, uranium doping has no effect on critical current. Preliminary SEM data suggested that the uranium is homogeneously distributed throughout the oxide core of the tape. The presence of 2212 and other secondary phases in the doped tapes suggest further refinement of the sintering procedure is necessary. (authors)

Low temperature irradiation effects on iron-boron based amorphous metallic alloys

International Nuclear Information System (INIS)

Audouard, Alain.

1983-01-01

Three iron-boron amorphous alloys and the crystalline Fe 3 B alloy have been irradiated at liquid hydrogen temperature. 2,4 MeV electron irradiation induces the creation of point defects in the amorphous alloys as well as in the crystalline Fe 3 B alloy. These point defects can be assimilated to iron ''Frenkel pairs''. They have been characterized by determining their intrinsic electrical resistivity and their formation volume. The displacement threshold energy of iron atoms has also been determined. 10 B fission fragments induce, in these amorphous alloys, displacement cascades which lead to stable vacancy rich zones. This irradiation also leads to a structural disorder in relation with the presence of defects. 235 U fission fragments irradiation modifies drastically the structure of the amorphous alloys. The results have been interpreted on the basis of the coexistence of two opposite processes which induce local disorder and crystallisation respectively [fr

Neutron irradiation test of copper alloy/stainless steel joint materials

International Nuclear Information System (INIS)

Yamada, Hirokazu; Kawamura, Hiroshi

2006-01-01

As a study about the joint technology of copper alloy and stainless steel for utilization as cooling piping in International Thermonuclear Experimental Reactor (ITER), Al 2 O 3 -dispersed strengthened copper or CuCrZr was jointed to stainless steel by three kinds of joint methods (casting joint, brazing joint and friction welding method) for the evaluation of the neutron irradiation effect on joints. A neutron irradiation test was performed to three types of joints and each copper alloy. The average value of fast neutron fluence in this irradiation test was about 2 x 10 24 n/m 2 (E>1 MeV), and the irradiation temperature was about 130degC. As post-irradiation examinations, tensile tests, hardness tests and observation of fracture surface after the tensile tests were performed. All type joints changed to be brittle by the neutron irradiation effect like each copper alloy material, and no particular neutron irradiation effect due to the effect of joint process was observed. On the casting and friction welding, hardness of copper alloy near the joint boundary changed to be lower than that of each copper alloy by the effect of joint procedure. However, tensile strength of joints was almost the same as that of each copper alloy before/after neutron irradiation. On the other hand, tensile strength of joints by brazing changed to be much lower than CuAl-25 base material by the effect of joint process before/after neutron irradiation. Results in this study showed that the friction welding method and the casting would be able to apply to the joint method of piping in ITER. This report is based on the final report of the ITER Engineering Design Activities (EDA). (author)

Geochemical correlations between uranium and other components in U-bearing formations of Ogcheon belt

International Nuclear Information System (INIS)

Lee, M.S.; Chon, H.T.

1980-01-01

Some components in uranium-bearing formations which consist mainly of black shale, slate and low grade coal-bearing formation of Ogcheon Belt were processed statistically in order to find out the geochemical correlations with uranium. Geochemical enrichment of uranium, vanadium and molybdenum in low grade coal-bearing formations and surrounding rocks is remarkable in the studied area. Geochemical correlation coefficient of uranium and molybdenum in the rocks displays about 0.6 and that of uranium and fixed carbon about 0.4. Uranium and vanadium in uranium-bearing low grade coals denote very high correlation with fixed carbon, which is considered to be responsible for enrichment of metallic elements, especially molybdenum. Close geochemical correlation of uranium-molybdenum couple in the rocks can be applied as a competent exploration guide to low grade uranium deposits of this area. (author)

Separation of 99Tc from irradiated molybdenum oxide by extraction with trioctilamine

International Nuclear Information System (INIS)

Carvalho, O.G. de.

1979-01-01

A separation method of sup( 99 m)Tc, from irradiated molybdenum oxide, by extraction with trioctilamino in 1,2 dichloroethane, 2% v/v is studied. Two preliminary studies are done: 1) Stablishment of the shaking time necessary to reach the equilibrium between the organic and the aqueous phase; 2) Choice of the concentration of solution of TOA in 1,2 dichloroethane to obtain the best separation conditions of sup( 99 m)Tc. After stablishing these two parameters, the study of extraction in solutions of hydrochloric nitric and sulfuric acids in different concentrations is done, followed by the study of the variation of extraction percentage of sup( 99 m)Tc in relation to the molybdenum oxide mass and the back-extraction of sup( 99 m)Tc to the aqueous phase with solutions of perchloric acid 1,0 and 0,1 N and ammonium hydroxide 1,0 N. (Author) [pt

Uranium determination in U-Al alloy with statistical tools support

International Nuclear Information System (INIS)

Furusawa, Helio Akira; Medalla, Felipe Quirino; Cotrim, Marycel Elena Barbosa; Pires, Maria Aparecida Faustino

2011-01-01

ICP-OES was used to quantify total uranium in natural UAl x powder alloy. A simple solubilisation procedure using diluted HNO 3 /HCl was successfully applied. Only 100 mg of sample were used which is an advantage over the volumetric methodologies. Only two dilutions were needed to reach measurable concentration. No other treatment was applied to the solutions. Calibration curves of three uranium lines (367.007, 385.958 and 409.014 nm) were evaluated using ANOVA. Comparing the indicators, the 367.007 nm line was the poorer one but exhibiting a R 2 = 0.998 and 0.9996 and 0.999 for the other two lines. No significant difference was found between these two lines. If needed, the 385.958 nm line could be used to quantify uranium in very low concentrations but with few advantages over the 409.014 nm line, if so. The average uranium concentration found was 0.80±0.01 μg.g-1, as expected for a predominant UAl 2 phase alloy. Higher uranium concentrations are also expected to be successfully quantified using these lines. In order to verify possibly inhomogeneity due to the high uranium concentration, one-way ANOVA was applied to 3 replicates. Homogeneity was confirmed measuring in both 385.958 and 409.014 nm lines. The uncertainty of solution homogeneity was estimated also in these two emission lines giving 0.006 and 0.005 μg.g-1, respectively. These two values are in compliance with the standard deviation of the average. (author)

Irradiation-induced microstructural changes in alloy X-750

International Nuclear Information System (INIS)

Kenik, E.A.

1997-01-01

Alloy X-750 is a nickel base alloy that is often used in nuclear power systems for it's excellent corrosion resistance and mechanical properties. The present study examines the microstructure and composition profiles in a heat of Alloy X-750 before and after neutron irradiation

Study of transformations by annealing of the body. Centred cubic {gamma} phase of uranium-molybdenum alloys; Etude des transformations par revenu de la phase {gamma} cubique centree des alliages uranium-molybdene

Energy Technology Data Exchange (ETDEWEB)

Mikailoff, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1959-06-15

By annealing at different temperatures, we have studied the transformations of the body centred cubic {gamma} phase for two alloys containing 6 and 10 per cent molybdenum by weight respectively. There is a return to the equilibrium state by formation of the stable {alpha} orthorhombic and {epsilon} ordered tetragonal phases, following two types of reaction: - pearlite transformation by nucleation and growth from the grain boundaries, preponderant when the annealing takes place at temperature above 400 deg. C, and identical for the two types of alloys. This reaction has already been studied by numerous authors, who have constructed the corresponding TTT curves, - transformation inside the grains of the quenched solid solution when annealing takes place at 400 deg. C or below: 6 per cent alloy - precipitation of fine a phase particles, followed by progressive ordering of the solid solution enriched in molybdenum, 10 per cent alloy - formation of small ordered regions and then a fine a phase precipitate. In the course of this work we have paid particular attention to the study of intragranular reactions after low-temperature annealing, the reactions involved in this case not having been explained up to the present. The {gamma} phase transformation has been studied by means of three techniques: micrography - microhardness tests - X-ray diffraction. (author) [French] Nous avons etudie les transformations par revenu a differentes temperatures, de la phase {gamma} cubique centree des alliages U-Mo trempes, pour deux alliages a 6 et a 10 pour cent de molybdene en poids. Il y a retour a l'etat d'equilibre par formation des phases stables {alpha} orthorhombique et quadratique ordonnee, suivant deux types de reactions: - transformation perlitique par germination et croissance a partir des joints de grains, preponderante lorsque le recuit a lieu a temperature superieure a 400 deg. C, et identique pour les deux types d'alliages. Cette reaction a deja ete etudiee par de nombreux

Development of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.

1997-01-01

The Reduced Enrichment Research and Test Reactor Program has continued its effort in the past 3 yr to develop use of low-enriched uranium (LEU) to produce the fission product 99 Mo. This work comprises both target and chemical processing development and demonstration. Two major target systems are now being used to produce 99 Mo with highly enriched uranium-one employing research reactor fuel technology (either uranium-aluminum alloy or uranium aluminide-aluminum dispersion) and the other using a thin deposit of UO 2 on the inside of a stainless steel (SST) tube. This paper summarizes progress in irradiation testing of targets based on LEU uranium metal foils. Several targets of this type have been irradiated in the Indonesian RSG-GAS reactor operating at 22.5 MW

Method for electrodeposition of nickel--chromium alloys and coating of uranium

International Nuclear Information System (INIS)

Stromatt, R.W.; Lundquist, J.R.

1975-01-01

High-quality electrodeposits of nickel-chromium binary alloys in which the percentage of chromium is controlled can be obtained by the addition of a complexing agent such as ethylenediaminetetraacetic disodium salt to the plating solution. The nickel-chromium alloys were found to provide an excellent hydrogen barrier for the protection of uranium fuel elements. (U.S.)

Mathematic modeling of reactor fuel radiation creep at example of uranium and its alloys

International Nuclear Information System (INIS)

Tarasov, V.A.

2001-01-01

The model of a radiation creep is explained within the framework of the mechanism of gliding and climbing dislocations based on the conception of a dislocation as not ideal sink for point radiation defects (PRD). The offered model is efficient for installed concentration PRD, considerably exceeding thermally steady state concentration. The gliding of dislocation are describing as due to moving dislocation kinks in Peierl's relief. The climbing of dislocation are describing as due to moving dislocation jogs. The mathematical model for the computer program simulating the offered model of radiation creep is developed. The complex of the computer programs simulating the radiation creep is developed. The computer simulation researches are conducted and the outcomes of a research of a kinetics of a flexible sliding and climbing dislocation interacting to obstacles of a various type (spherical centre of extension, dislocation prismatic loop and their spatially random distributions) for various installed concentration PRD, external loadings and temperatures are represented. The curves of installed rate of a radiation creep from temperature for uranium and its alloys with small additions of molybdenum (from 0,9 to 1,3 %) are obtained

Production of fusion radionuclides: Molybdenum-99/ Iodine - 131 and Xenon-133; Produccion de los radionucleidos de fision: Molibdeno-99, Yodo-131 y Xenon-133

Energy Technology Data Exchange (ETDEWEB)

Barrachina, M; Carrillo, D

1982-07-01

This report presents a new radiochemical method for industrial production of the radionuclides: molybdenum-99, iodine-131 and xenon-133. The above mentioned method based on the alkaline metathesis reaction of irradiated uranium (IV) fluoride, presents the best characteristics for the proposed objective. The study deals with the analysis of that reaction and the separation and purification processes. (Author) 71 refs.

GRAIN-BOUNDARY PRECIPITATION UNDER IRRADIATION IN DILUTE BINARY ALLOYS

Institute of Scientific and Technical Information of China (English)

S.H. Song; Z.X. Yuan; J. Liu; R.G.Faulkner

2003-01-01

Irradiation-induced grain boundary segregation of solute atoms frequently bring about grain boundary precipitation of a second phase because of its making the solubility limit of the solute surpassed at grain boundaries. Until now the kinetic models for irradiation-induced grain boundary precipitation have been sparse. For this reason, we have theoretically treated grain boundary precipitation under irradiation in dilute binary alloys. Predictions ofγ'-Ni3Si precipitation at grain boundaries ave made for a dilute Ni-Si alloy subjected to irradiation. It is demonstrated that grain boundary silicon segregation under irradiation may lead to grain boundaryγ'-Ni3 Si precipitation over a certain temperature range.

Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

Energy Technology Data Exchange (ETDEWEB)

Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Baek, J. S [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.

2014-04-30

It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

In-situ leaching of crownpoint, NM, uranium ore: Part 7 - Laboratory study of chemical agents for molybdenum restoration

International Nuclear Information System (INIS)

Strom, E.T.; Vogt, T.C.

1987-01-01

One possible drawback to the use of an in-situ leaching to recover uranium is the potential release of previously insoluble chemical species into the formation water. Before a pilot test of in-situ uranium leaching at Crownpoint, NM, was begun, extensive laboratory studies were undertaken to develop chemical methods for treating one possible contaminant, molybdenum (Mo). New Mexico regulations restrict the amount of Mo permissable in formation waters after leaching to less than 1 ppm. Two techniques to restore Mo after leaching were studied with core and pack tests. These studies suggest that if Mo restoration problems occur in the field, the use of precipitating agents such as Ca/sup 2+/ or reducing agents such as Fe/sup 2+/ may be helpful in ameliorating such problems

Evaluation of Ion Irradiation Behavior of ODS Alloys

Energy Technology Data Exchange (ETDEWEB)

Jang, Jin Sung; Kim, Min Chul; Hong, Jun Hwa; Han, Chang Hee; Chang, Young Mun; Bae, Chang Soo; Bae, Yoon Young; Chang, Moon Hee

2006-08-15

FM steel (Grade 92) and ODS alloy(MA956) specimens were ion irradiated with 122 MeV Ne ions. Irradiation temperatures were about 450 and 550 .deg. C and the peak dose was 1, 5, and 10 dpa. Cross-sectional TEM samples were prepared by the electrolytic Ni-plating after pre-treatment of the irradiated specimens. Irradiation cavities in FM steel and ODS alloy specimens were not much different in size; about 20 nm in diameter in both specimens irradiated at around 450 .deg. C. However, the size distribution of cavities in FM steel specimens was broader than that in ODS alloy specimen, indicating that the cavity growth probably via coalescence). It was noticeable that the location and the preferential growth of the cavities in FM steel specimens: cavities on the PAGB (prior austenite grain boundary) was significantly larger than those within the grains. This could be an important issue for the mechanical properties, especially high temperature creep, fracture toughness, and so on. The dependency of the dose threshold and swelling on the ratio of the inert gas concentration/dpa was analysed for the various irradiation source, including He, Ne, Fe/He, and fast neutron, and the empirical correlation was established.

Evaluation of Ion Irradiation Behavior of ODS Alloys

International Nuclear Information System (INIS)

Jang, Jin Sung; Kim, Min Chul; Hong, Jun Hwa; Han, Chang Hee; Chang, Young Mun; Bae, Chang Soo; Bae, Yoon Young; Chang, Moon Hee

2006-08-01

FM steel (Grade 92) and ODS alloy(MA956) specimens were ion irradiated with 122 MeV Ne ions. Irradiation temperatures were about 450 and 550 .deg. C and the peak dose was 1, 5, and 10 dpa. Cross-sectional TEM samples were prepared by the electrolytic Ni-plating after pre-treatment of the irradiated specimens. Irradiation cavities in FM steel and ODS alloy specimens were not much different in size; about 20 nm in diameter in both specimens irradiated at around 450 .deg. C. However, the size distribution of cavities in FM steel specimens was broader than that in ODS alloy specimen, indicating that the cavity growth probably via coalescence). It was noticeable that the location and the preferential growth of the cavities in FM steel specimens: cavities on the PAGB (prior austenite grain boundary) was significantly larger than those within the grains. This could be an important issue for the mechanical properties, especially high temperature creep, fracture toughness, and so on. The dependency of the dose threshold and swelling on the ratio of the inert gas concentration/dpa was analysed for the various irradiation source, including He, Ne, Fe/He, and fast neutron, and the empirical correlation was established

Effect of irradiation on the tensile properties of niobium-base alloys

International Nuclear Information System (INIS)

Grossbeck, M.L.; Heestand, R.L.; Atkin, S.D.

1986-11-01

The alloys Nb-1Zr and PWC-11 (Nb-1Zr-0.1C) were selected as prime candidate alloys for the SP-100 reactor. Since the mechanical properties of niobium alloys irradiated to end-of-life exposure levels of about 2 x 10 26 neutrons/m 2 (E > 0.1 MeV) at temperatures above 1300 K were not available, an irradiation experiment (B-350) in EBR-II was conducted. Irradiation creep, impact properties, bending fatigue, and tensile properties were investigated; however, only tensile properties will be reported in this paper. The tensile properties were studied since they easily reveal the common irradiation phenomena of hardening and embrittlement. Most attention was directed to testing at the irradiation temperature. Further testing was conducted at lower temperatures in order to scope the behavior of the alloys in cooldown conditions

Structure and properties of UV-irradiated LLDPE and alloy of PA66 and the irradiated LLDPE

International Nuclear Information System (INIS)

Ran Qianping; Zou Hua; Wu Shishan; Shen Jian

2006-01-01

Some oxygen-containing groups such as C=O, C-O and -OH were introduced onto linear low-density polyethylene (LLDPE) chains by UV irradiation in air. Their concentration increased with the irradiation time. Crystal shape of the irradiated LLDPE remained an orthorhombic structure, while space of the crystalline plane kept unchanged. The melting temperature and crystallinity decreased due to the LLDPE chain scission into small molecules compound and crystalline defects caused by UV irradiation. Compared with pristine LLDPE, hydrophilicity of the irradiated LLDPE increased due to the introduction of polar oxygen-containing groups, but the tensile strength decreased due to the LLDPE chain degradation and reduction of crystallinity. The temperature of initial weight loss of the irradiated LLDPE was lower than that of pristine LLDPE. An alloy of PA66 and the irradiated LLDPE (irradiated PA66/LLDPE) was prepared by melting blend at 260-270 degree C. Compared to non-irradiated PA66/LLDPE alloy, dispersion of LLDPE particles in the irradiated PA66/LLDPE alloy and interfacial interactions between the components were markedly improved. Therefore, tensile strength and impact strength of the irradiated PA66/ LLDPE were higher than those of the control. (authors)

X-ray topography of uranium alloys; Topographie aux rayons X d'alliages d'uranium

Energy Technology Data Exchange (ETDEWEB)

Le Naour, L [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

A description of the structure of uranium alloys has been made using the data obtained by X-ray diffraction techniques derived from the Berg-Barrette method. In the first.stage the use of a monochromatic beam of X-rays having a very low divergence makes it possible to obtain very reproducible and exact numerical data concerning the grain and sub-grain sizes, and also the distribution of the sizes. It is thereby possible to detect any disorientation greater than 30 seconds of arc.The results obtained have been completed using a variable incidence device which- gives simultaneously an overall picture of a grain and an idea of the importance of internal disorientations; a more rigorous measurement of this latter parameter is then deduced from the Debye-Scherrer diagrams obtained using a fine-focus equipment. Observations are carried out on various one-phase or two phase uranium alloys which are compared successively to technical and to high-purity uranium. It is shown that the use of X-ray topographies, although limited in certain respects, allows a quantitative characterization of the structure. (author) [French] Une description des structures d'alliages d'uranium a ete faite a partir des donnees fournies par des techniques de diffraction de rayons X derivees de la methode de BERG--BARRETT. Dans une premiere etape, l'utilisation d'un faisceau de rayons X monochromatique et de tres faible divergence permet d'obtenir des donnees numeriques precises et tres reproductibles, relatives aux dimensions des grains, des sous-grains et a la distribution de ces grandeurs. Toute desorientation superieure a 30 secondes d'arc peut ainsi etre decelee. Les resultats obtenus ont ete completes en utilisant un montage a incidence variable, qui fournit simultanement l'image globale d'un grain et l'ordre de grandeur des desorientations internes; une mesure plus rigoureuse de ce dernier parametre se deduit ensuite de diagrammes DEBYE SHERRER realises avec un montage a foyer fin. Des

Vapor corrosion of aluminum cladding alloys and aluminum-uranium fuel materials in storage environments

International Nuclear Information System (INIS)

Lam, P.; Sindelar, R.L.; Peacock, H.B. Jr.

1997-04-01

An experimental investigation of the effects of vapor environments on the corrosion of aluminum spent nuclear fuel (A1 SNF) has been performed. Aluminum cladding alloys and aluminum-uranium fuel alloys have been exposed to environments of air/water vapor/ionizing radiation and characterized for applications to degradation mode analysis for interim dry and repository storage systems. Models have been developed to allow predictions of the corrosion response under conditions of unlimited corrodant species. Threshold levels of water vapor under which corrosion does not occur have been identified through tests under conditions of limited corrodant species. Coupons of aluminum 1100, 5052, and 6061, the US equivalent of cladding alloys used to manufacture foreign research reactor fuels, and several aluminum-uranium alloys (aluminum-10, 18, and 33 wt% uranium) were exposed to various controlled vapor environments in air within the following ranges of conditions: Temperature -- 80 to 200 C; Relative Humidity -- 0 to 100% using atmospheric condensate water and using added nitric acid to simulate radiolysis effects; and Gamma Radiation -- none and 1.8 x 10 6 R/hr. The results of this work are part of the body of information needed for understanding the degradation of the A1 SNF waste form in a direct disposal system in the federal repository. It will provide the basis for data input to the ongoing performance assessment and criticality safety analyses. Additional testing of uranium-aluminum fuel materials at uranium contents typical of high enriched and low enriched fuels is being initiated to provide the data needed for the development of empirical models

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

Science.gov (United States)

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

1975-01-01

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

Determination of irradiated uranium in far-field contaminated areas of Belarus

International Nuclear Information System (INIS)

Mironov, V.; Pribylev, S.; Hotchkis, M.; Child, D.

2006-01-01

The possibility of using U 236 as an indicator for irradiated uranium is shown. The sensitivity of AMS is high enough for measurements of 236 U/ 238 U ratios down to 10 -9 on micrograms of uranium and therefore for the detection of Chernobyl originated uranium in the remote regions of radioactive fallout. (authors)

Microstructural and microchemical evolution in vanadium alloys by heavy ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Sekimura, Naoto; Kakiuchi, Hironori; Shirao, Yasuyuki; Iwai, Takeo [Tokyo Univ. (Japan)

1996-10-01

Microstructural and microchemical evolution in vanadium alloys were investigated using heavy ion irradiation. No cavities were observed in V-5Cr-5Ti alloys irradiated to 30 dpa at 520 and 600degC. Energy dispersive X-ray spectroscopy analyses showed that Ti peaks around grain boundaries. Segregation of Cr atoms was not clearly detected. Co-implanted helium was also found to enhance dislocation evolution in V-5Cr-5Ti. High density of matrix cavities were observed in V-5Fe alloys irradiated with dual ions, whereas cavities were formed only around grain boundaries in single ion irradiated V-5Fe. (author)

Neutronic and thermal hydraulic analyses of LEU targets irradiated in a research reactor for Molybdenum-99 production

International Nuclear Information System (INIS)

Jo, Daeseong; Lee, Kyung-Hoon; Kim, Hong-Chul; Chae, Heetaek

2014-01-01

Highlights: • Neutronic and thermal hydraulic analyses of irradiated fuel plates for Molybdenum-99. • Heat production during and after irradiation was evaluated using MCNP and ORIGEN-APR. • Cooling capacities under various cooling conditions were evaluated using TMAP. • Natural convective cooling was adequate for the decay power after 0.03 h from withdrawal. • Maximum temperature of the target decayed for 24 h does not exceed the blistering threshold. - Abstract: Neutronic and thermal hydraulic analyses of irradiated fuel plates for Molybdenum-99 production in a research reactor were performed to investigate (1) the heat production during irradiation, (2) decay heat after irradiation, and (3) cooling capacities under various cooling conditions. The heat production on the target plates irradiated in the core was evaluated using the MCNP code. The decay heat after irradiation was evaluated using the ORIGEN-APR code, and compared against ANSI/ANS-5.1-1979. The cooling capacities of forced convective cooling during irradiation and natural convective cooling after irradiation were estimated using the TMAP code. An equilibrium core with different core statuses i.e., BOC, MOC, and EOC was used to evaluate power released from the targets and the axial power distribution. Based on the neutronic calculations, thermal margins i.e., the maximum wall temperature, minimum ONB temperature margin, and minimum CHF ratio were estimated, and the cooling strategy of the fission Mo targets was discussed. The targets were cooled by forced convective cooling during irradiation, and cooled by natural convective cooling after irradiation. For a further production process, the targets transported to a hot cell were exposed to the air, and cooled by natural convection cooling in air. As a result, the maximum wall temperature remained below the ONB temperature while the targets were under water, and the maximum wall temperature remained under the blistering limit while the targets

Effect of passivation with CO on the electrochemical corrosion behavior of uranium-niobium alloy

International Nuclear Information System (INIS)

Fu Xiaoguo; Dai Lianxin; Zou Juesheng; Bai Chaomao; Wang Xiaolin

2000-01-01

Electrochemical studies are performed to investigate the corrosion resistance of uranium-niobium alloy before and after passivated with carbon monoxide. Using X-ray photoelectron spectroscopy (XPS), the surface composition of specimen passivated with carbon monoxide is determined. The corrosion resistance of uranium-niobium alloy is well improved because the passive layer (UC/UC x O y + Nb 2 O 5 + UO 2 ) on surface serves as passive film and increases the anodic impedance after the specimen is passivated with carbon monoxide

Irradiation effects of the zirconium oxidation and the uranium diffusion in zirconia; Effets d'irradiation sur l'oxydation du zirconium et la diffusion de l'uranium dans la zircone

Energy Technology Data Exchange (ETDEWEB)

Bererd, N

2003-07-01

The context of this study is the direct storage of spent fuel assemblies after operation in reactor. In order to obtain data on the capacities of the can as the uranium diffusion barrier, a fundamental study has been carried out for modelling the internal cladding surface under and without irradiation. The behaviour of zirconium in reactor conditions has at first been studied. A thin uranium target enriched with fissile isotope has been put on a zirconium sample, the set being irradiated by a thermal neutrons flux leading to the fission of the deposited uranium. The energetic history of the formed fission products has revealed two steps: 1)the zirconium oxidation and 2)the diffusion of uranium in the zirconia formed at 480 degrees C. A diffusion coefficient under irradiation has been measured. Its value is 10{sup -15} cm{sup 2}.s{sup -1}. In order to be able to reveal clearly the effect of the irradiation by the fission products on the zirconium oxidation, measurements of thermal oxidation and under {sup 129}Xe irradiation have been carried out. They have shown that the oxidation is strongly accelerated by the irradiation and that the temperature is negligible until 480 degrees C. On the other hand, the thermal diffusion of the uranium in zirconium and in zirconia has been studied by coupling ion implantation and Rutherford backscattering spectroscopy. This study shows that the uranium diffuses in zirconium and is trapped in zirconia in a UO{sub 3} shape. (O.M.)

In-situ leaching of Crownpoint, New Mexico, uranium ore: Part 7 - laboratory study of chemical agents for molybdenum restoration

International Nuclear Information System (INIS)

Strom, E.T.; Vogt, T.C.

1985-01-01

While in-situ leaching has significant advantages over conventional uranium recovery methods, one possible drawback to its use is the potential release of previously insoluble chemical species into the formation water. Before Mobil began a pilot test of in-situ uranium leaching at Crownpoint, New Mexico, extensive laboratory studies were undertaken to develop chemical methods for treating one possible contaminant, molybdenum (Mo). In-situ production of uranium entails oxidizing uranium from the insoluble +4 oxidation state to the soluble, readily complexed +6 state. However, this process also transforms insoluble Mo +4 compounds such as molybdenite or jordesite, MoS 2 , into the soluble T6 form, molybdate, Mo0 4 2- . New Mexico regulations restrict the amount of Mo permissible in formation waters after leaching to less than one ppm. Conceptually, Mo restoration after leaching can be dealt with in one of two ways. (1) The oxidizing environment can be left unchanged with something added to render the molybdate ion insoluble or (2) the environment can be changed to a reducing one, converting the Mo back to the less soluble +4 oxidation state

A Model for High-Strain-Rate Deformation of Uranium-Niobium Alloys

Energy Technology Data Exchange (ETDEWEB)

F.L.Addessio; Q.H.Zuo; T.A.Mason; L.C.Brinson

2003-05-01

A thermodynamic approach is used to develop a framework for modeling uranium-niobium alloys under the conditions of high strain rate. Using this framework, a three-dimensional phenomenological model, which includes nonlinear elasticity (equation of state), phase transformation, crystal reorientation, rate-dependent plasticity, and porosity growth is presented. An implicit numerical technique is used to solve the evolution equations for the material state. Comparisons are made between the model and data for low-strain-rate loading and unloading as well as for heating and cooling experiments. Comparisons of the model and data also are made for low- and high-strain-rate uniaxial stress and uniaxial strain experiments. A uranium-6 weight percent niobium alloy is used in the comparisons of model and experiment.

Solute segregation and void formation in ion-irradiated vanadium-base alloys

International Nuclear Information System (INIS)

Loomis, B.A.; Smith, D.L.

1985-01-01

The radiation-induced segregation of solute atoms in the V-15Cr-5Ti alloys was determined after either single- dual-, or helium implantation followed by single-ion irradiation at 725 0 C to radiation damage levels ranging from 103 to 169 dpa. Also, the effect of irradiation temperature (600-750 0 C) on the microstructure in the V-15Cr-5Ti alloy was determined after single-ion irradiation to 200 and 300 dpa. The solute segregation results for the single- and dual-ion irradiated alloy showed that the simultaneous production of irradiation damage and deposition of helium resulted in enhanced depletion of Cr solute and enrichment of Ti, C and S solute in the near-surface layers of irradiated specimens. The observations of the irradiation-damaged microstructures in V-15Cr-5Ti specimens showed an absence of voids for irradiations of the alloy at 600-750 0 C to 200 dpa and at 725 0 C to 300 dpa. The principle effect on the microstructure of these irradiations was to induce the formation of a high density of disc-like precipitates in the vicinity of grain boundaries and intrinsic precipitates and on the dislocation structure. 8 references, 4 figures

Mechanical properties of molybdenum-titanium alloys micro-structurally controlled by multi-step internal nitriding

International Nuclear Information System (INIS)

Nagae, M.; Yoshio, T.; Takemoto, Y.; Takada, J.; Hiraoka, Y.

2001-01-01

Internally nitrided dilute Mo-Ti alloys having a heavily deformed microstructure near the specimen surface were prepared by a novel two-step nitriding process at 1173 to 1773 K in N 2 gas. For the nitrided specimens three-point bend tests were performed at temperatures from 77 to 298 K in order to investigate the effect of microstructure control by internal nitriding on the ductile-to-brittle transition temperature (DBTT) of the alloy Yield strength obtained at 243 K of the specimen maintaining the deformed microstructure by the two-step nitriding was about 1.7 times as much as recrystallized specimen. The specimen subjected to the two-step nitriding was bent more than 90 degree at 243 K, whereas recrystallized specimen was fractured after showing a slight ductility at 243 K. DBTT of the specimen subjected to the two-step nitriding and recrystallized specimen was about 153 K and 203 K, respectively. These results indicate that multi-step internal nitriding is very effective to the improvement in the embrittlement by the recrystallization of molybdenum alloys. (author)

Relationship of microstructure and tensile properties for neutron-irradiated vanadium alloys

International Nuclear Information System (INIS)

Loomis, B.A.; Smith, D.L.

1990-01-01

The microstructures in V-15Cr-5Ti, V-10Cr-5RTi, V-3Ti-1Si, V-15Ti-7.5Cr, and V-20Ti alloys were examined by transmission electron microscopy after neutron irradiation at 600 degree C to 21--84 atom displacements per atom in the Materials Open Test Assembly of the Fast Flux Test Facility. The microstructures in these irradiated alloys were analyzed to determine the radiation-produced dislocation density, precipitate number density and size, and void number density and size. The results of these analyses were used to compute increases in yield stress and swelling of the irradiated alloys. The computed increase in yield stress was compared with the increase in yield stress determined from tensile tests on these irradiated alloys. This comparison made it possible to evaluate the influence of alloy composition on the evolution of radiation-damaged microstructures and the resulting tensile properties. 11 refs

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

International Nuclear Information System (INIS)

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

2015-01-01

In this paper, we report on the temperature dependent surface modifications in molybdenum (Mo) samples due to 100 eV He + ion irradiation in extreme conditions as a potential candidate to plasma-facing components in fusion devices alternative to tungsten. The Mo samples were irradiated at normal incidence, using an ion fluence of 2.6 × 10 24 ions m −2 (with a flux of 7.2 × 10 20 ions m −2 s −1 ). Surface modifications have been studied using high-resolution field emission scanning electron-(SEM) and atomic force (AFM) microscopy. At 773 K target temperature homogeneous evolution of molybdenum nanograins on the entire Mo surface were observed. However, at 823 K target temperature appearance of nano-pores and pin-holes nearby the grain boundaries, and Mo fuzz in patches were observed. The fuzz density increases significantly with target temperatures and continued until 973 K. However, at target temperatures beyond 973 K, counterintuitively, a sequential reduction in the fuzz density has been seen till 1073 K temperatures. At 1173 K and above temperatures, only molybdenum nano structures were observed. Our temperature dependent studies confirm a clear temperature widow, 823–1073 K, for Mo fuzz formation. Ex-situ high resolution X-ray photoelectron spectroscopy studies on Mo fuzzy samples show the evidence of MoO 3 3d doublets. This elucidates that almost all the Mo fuzz were oxidized during open air exposure and are thick enough as well. Likewise the microscopy studies, the optical reflectivity measurements also show a sequential reduction in the reflectivity values (i.e., enhancement in the fuzz density) up to 973 K and after then a sequential enhancement in the reflectivity values (i.e., reduction in the fuzz density) with target temperatures. This is in well agreement with microscopy studies where we observed clear temperature window for Mo fuzz growth

Post-irradiation annealing of coarse-grained model alloys

Energy Technology Data Exchange (ETDEWEB)

Ray, P H.N.; Wilson, C; McElroy, R J [AEA Reactor Services, Harwell (United Kingdom)

1994-12-31

Thermal ageing and irradiation studies have been carried out on three model alloys (JPC, JPB, JPG) that have identical compositions except for different levels of phosphorus and/or copper. They have been irradiated in three conditions, as-received, heat treated to produce a coarse grained microstructure (similar to heat-affected-zone), and in this condition further aged at 450 C to produce a temper embrittled condition. One of the alloy have been subject to a post-irradiation anneal. The effect of these treatments on mechanical property changes has been characterized by Charpy testing and Vickers hardness measurements; the phosphorus segregation has been studied by a combination of STEM and Auger techniques.

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

Science.gov (United States)

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

1983-10-01

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

Metallurgical structures in a high uranium-silicon alloy

International Nuclear Information System (INIS)

Wyatt, B.S.; Berthiaume, L.C.; Conversi, J.L.

1968-10-01

The effects of fabrication and heat treatment variables on the structure of a uranium -- 3.96 wt% silicon alloy have been studied using optical microscopy, quantitative metallography and hardness determinations. It has been shown that an optimum temperature exists below the peritectoid temperature where the maximum amount of transformation to U 3 Si occurs in a given period of time. The time required to fully transform an as-cast alloy at this optimum temperature is affected by the size of the primary U 3 Si 2 dendrites. With a U 3 Si 2 particle size of <12 μm complete transformation can be achieved in four hours. (author)

Neutron irradiation effect on the strength of jointed Ti-6Al-4V alloy

International Nuclear Information System (INIS)

Ishiyama, Shintaro; Miya, Naoyuki

2002-01-01

In order to investigate applicability of Ti alloy to large scaled structural material for fusion reactors, irradiation effect on the mechanical properties of Ti-6Al-4V alloy and its TIG welded material was investigated after neutron irradiation (temperature: 746-788K, fluence: 2.8 x 10 23 n/m 2 (>0.18 MeV). The following results were obtained. (1) Irradiated Ti alloy shows about 20-30% increase of its tensile strength and large degradation of fracture elongation, comparing with those of unirradiated Ti alloy. (2) TIG welded material behaves as Ti alloy in its tensile test, however, shows 30% increase of area reduction in 373-473K, whereas 1/2 degradation of area reduction over 600K. (3) Irradiated TIG welded material behaves heavier embrittlement than that of irradiated Ti alloy. (4) Charpy impact properties of un- and irradiated Ti alloys shift to ductile from brittle fracture and transition temperature shift, ΔT was estimated as about 100K. (5) Remarkable increase of hardness was found, especially in HAZ of TIG welded material after irradiation. (author)

A method for the electrolytic coating of uranium or uranium alloy parts, and parts thus obtained

International Nuclear Information System (INIS)

1973-01-01

A method, preceded by a surface treatment, for applying an electrolytic coating (e.g. of nickel) on uranium, or uranium alloy parts. This method is characterized in that the previous surface treatment comprises a chemical removal of grease in halogenated solvent bath (free from halogen ions) and an anodic scouring in a buffered aqueous solution solution of an acid free from halogen ions. The coating can be applied to fuel elements for nuclear industry, counter-weight for aeronautics and space industries and to radiation shields [fr

Low content uranium alloys for nuclear fuels

International Nuclear Information System (INIS)

Aubert, H.; Laniesse, J.

1964-01-01

A description is given of the structure and the properties of low content alloys containing from 0.1 to 0.5 per cent by weight of Al, Fe, Cr, Si, Mo or a combination of these elements. A study of the kinetics and of the mode of transformation has made it possible to choose the most satisfactory thermal treatment. An attempt has been made to prepare alloys suitable for an economical industrial development having a small α grain structure without marked preferential orientation, with very fine and stable precipitates as well as a high creep-resistance. The physical properties and the mechanical strength of these alloys are given for temperatures of 20 to 600 deg C. These alloys proved very satisfactory when irradiated in the form of normal size fuel elements. (authors) [fr

Low activation ferritic alloys

Science.gov (United States)

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.

Contribution towards the study of β→α transformation in uranium and its alloys (1962)

International Nuclear Information System (INIS)

Aubert, H.

1962-05-01

The kinetics of the transformation of uranium alloys containing 0.5 - 0.75 - 1.0 - 1.5 and 3 atoms per cent have been studied. The influence of heat treatment before decomposition has been discussed. The study of the transformation characteristics such as kinetics, residual phases, phenomena connected with the coherence between phases, reversibility below the equilibrium temperature, shows the following mechanisms exhibited during the decomposition of the β phase on lowering the temperature: 1 ) eutectoid, 2) bainitic, 3) martensitic. The study of the TTT diagrams of alloys containing decreasing percentages of chromium indicates that the unalloyed uranium transforms without maintaining the coherence above 600 deg. C, where as at lower temperatures the transformation is mainly martensitic. The various alloying elements can be characterised by their influence on the three TTT curves corresponding to the three possible transformation mechanisms. The ability of the uranium alloys to alpha grain refining during isothermal decomposition or ambient temperature quenching is directly connected with the characteristics of the TTT diagrams and especially to the mode of bainitic transformation. (author) [fr

The reduction of organic component content, of molybdenum and coloidal silica in chlorosodium uraniferous eluates

International Nuclear Information System (INIS)

Filip, Gheorghe; Panturu, Eugenia; Radulescu, Rozalia; Predescu, Cristian; Filip, Dorin Gheorghe

2006-01-01

In this paper, the experimental results obtained at a laboratory facility for molybdenum, silica and organic substances removal from sodium chloride uraniferous eluate solutions are presented. The aim of this study is to obtain uranium yellow cakes having product quality specification, in order to enable their conversion in nuclear grade uranium dioxide. The studied variables were the contents for silica and organic substances at the acidifying and filtration of uranium eluates, residual molybdenum content in refined eluates after adsorption on activated carbon MN+P type. The chemical characterization of uranium concentrate obtained from the refined eluate is presented. (author)

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

Science.gov (United States)

Yamakawa, K.; Shimomura, Y.

1999-01-01

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

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

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