Mixture Of Uranium Research Articles

Determination of 236U/238U isotope ratio in contaminated environmental samples using different ICP-MS instruments

This paper considers use of the 236U isotope to monitor the spent uranium from nuclear fallout using inductively coupled plasma mass spectrometry (ICP-MS) in soil samples collected in the vicinity of the Chernobyl Nuclear Power Plant (NPP). Sector field ICP-MS (ICP-SFMS) and quadrupole based ICP-MS without and with hexapole collision cell (ICP-CC-MS) were used for uranium isotope analysis. In addition, a multiple ion collector ICP-MS (MC-ICP-MS) was used for high-precision isotope ratio measurements. A 238U+ ion sensitivity of 18 GHz ppm−1, 12.4 GHz ppm−1 and 16 GHz ppm−1, respectively, was observed in ICP-SFMS, ICP-CC-MS and MC-ICP-MS with ultrasonic nebulizer. An absolute sensitivity of 3600 counts fg−1 was achieved for uranium by using a direct injection high efficiency nebulizer for solution introduction in ICP-SFMS. The detection limit for 236U was in the fg g−1 range and abundance ratio sensitivity for 236U/238U was 5 × 10−6, 3 × 10−7, 6 × 10−7, and less than 3 × 10−7 in ICP-SFMS, MC-ICP-MS, quadrupole ICP-MS and ICP-CC-MS at mass resolution m/Δm = 300, respectively. Interlaboratory comparison yielded a good accuracy (0.4–1.6%) of 236U/238U isotope ratios ranging from 1.5 × 10−3 to 3.2 × 10−4 measured in samples containing 100 ng of uranium. The 236U/238U isotope ratios and spent uranium fraction were determined in Chernobyl soil samples using single ion detector ICP-MS and multiple ion collector ICP-MS (MC-ICP-MS). In comparison to 235U/238U isotope ratio, the 236U/238U isotope ratio provided more sensitive and accurate determination of the portion of spent uranium from Chernobyl NPP in spent/natural uranium mixture in soil samples down to 0.1%. The concentration of Chernobyl spent uranium in upper 0–10 cm soil layers in investigated areas in the vicinity of Chernobyl NPP amounts to 2.4 × 10−9 g g−1 to 8.1 × 10−7 g g−1 depending mainly on the distance to the Chernobyl reactor.

Isotopic analysis of uranium and plutonium using ICP-MS and estimation of burn-up of spent uranium in contaminated environmental samples

Environmental monitoring of actinides and evaluation of the contamination source (nuclear weapons tests, nuclear power plant and nuclear reprocessing plant accidents, etc.) requires precise and accurate isotopic analysis of actinides, especially uranium and plutonium. Double-focusing sector-field inductively coupled plasma mass spectrometry (ICP-SFMS) using a low-flow microconcentric nebulizer with membrane desolvation, “Aridus”, was applied for isotopic measurements of uranium and plutonium at the ultratrace level. The detection limit (3σ) for 236U and 239Pu after chemical extraction was 0.2 pg l−1 in aqueous solution and 0.04 pg g−1 in soil, respectively. 235U/238U, 236U/238U and 240Pu/239Pu isotope ratios were measured in soil samples collected within the 30 km zone around the Chernobyl nuclear power plant. The average 240Pu/239Pu isotope ratio in contaminated surface soil was 0.396 ± 0.014. The burn-up grade and the portion of spent uranium in the spent uranium/natural uranium mixture in soil were calculated using the iteration method. A slight variation in the burn-up grade of spent reactor uranium was revealed by analyzing 235U/238U and 236U/238U isotope ratios. A relationship between the 240Pu/239Pu isotope ratio and burn-up of spent uranium was observed.

New determination of the237Thhalf-life

Nuclide ${}^{237}\mathrm{Th}$ has been produced via multinucleon transfer reactions by 60 MeV/nucleon ${}^{18}\mathrm{O}$ ion irradiation of natural uranium. Thorium was radiochemically separated from the mixture of uranium and reaction products. The activity of thorium was measured by using a HPGe detector. The ${}^{237}\mathrm{Th}$ has been identified by measuring the growth and decay of the $\ensuremath{\gamma}$ rays from its daughter ${}^{237}\mathrm{Pa}.$ The half-life of ${}^{237}\mathrm{Th}$ was determined to be $4.69\ifmmode\pm\else\textpm\fi{}0.60$ min. Not only is the new value in agreement with the earlier one of $5.0\ifmmode\pm\else\textpm\fi{}0.9$ min obtained in our previous work, but also its experimental precision was improved.

Positive reactivity effect in hollow critical assemblies with moderated neutrons

Computational results obtained using the neutron-gas-dynamic codes for fission bursts in supercritical hollow assemblies of fissioning substances with a moderated neutron spectrum, specifically, in assemblies containing a neutron absorber, with a core consisting of a mixture of uranium, enriched to 90% with235U, and polyethylene are presented. Sharp manifestations of positive reactivity in the assemblies and a difference of the effect from that in fast systems are found. It is shown that in assemblies with moderated neutrons, as compared with fast systems, 1) the scale of the increase in the energy release in a burst is substantially reduced by inserting boron into the cavity; 2) in many cases the effect has the classical form with real growth of keff and λ in time; 3) the effect is essentially independent of the properties of the moderator in the cavity; and 4) the effect is less dangerous with respect to nuclear safety. 2 figures, 3 tables, 12 references.

Scientific and Technical Problems of the Nuclear Fuel Cycle

and the need to ensure an acceptable nuclear fuel cycle for future generations of nuclear power plants requires a comprehensive and systematic approach. The peculiar way in which nuclear power came into existence has not made it impossible to develop such an approach for handling spent fuel, and this is the main reason for the problems, which have now accumulated in the fuel cycle. Providing Fissioning Materials for Nuclear Power. The effect of stockpiled reserves and the quantity and quality of wastes on the strategy for selecting fissioning materials is not entirely clear. The forms and types of uranium compounds (oxide, metal, nitride, carbide) and plutonium compounds (oxide, metal, other compounds) which are used for fabricating fuel need to be optimized, and substantiation is needed for the possibility of using thorium (Table 1). The entire list of special raw materials and fissioning materials, which are present in the enterprises of the Ministry of Atomic Energy, includes 127 names. The optimization criteria for choosing various types of fuel and the time periods for putting them into use need to be determined. This requires the development of an integrated database. The priorities in choosing the materials are as follows: – nuclear and radiation safety in manufacturing and handling work (storage, shipment); – maximum reactor safety; – possibility of implementing an efficient technology for handling spent fuel, reducing to a minimum the volumes of wastes and radiation effects on the environment; – fuel-cycle economics; – remaining life of raw material resources (domestic and foreign) and the accessibility of these resources. Optimization calculations substantiating the choice of fissioning materials must take account of these priorities. Constructions of Reactor Fuel Elements and Fuel Assemblies. The structural materials, dimensions, shape, special features of the construction, the burnup specifications, and other properties must take account of the economic aspects of the operation of a nuclear power plant and the handling of spent fuel and wastes (Table 2). The main forms of fissioning materials used in nuclear fuel are as follows: – uranium oxides with different degrees of enrichment; the present cost of uranium obtained by additional enrichment of the hexafluoride from wastes containing 0.25% U, is much lower than in mined uranium; – uranium metal (ADE reactors), uranyl-nitrate alloy, oxides; – mixture of uranium and plutonium oxides (mixed uranium–plutonium fuel); – nitrides (uranium, plutonium, or mixed) obtained from metals or reduced from oxides. The main criteria for selecting the type of fuel are: – the cost of fresh fuel on the domestic and world markets; – the existence of an economically profitable infrastructure for handling such fuel at all stages of its life; – quality, satisfying the customer, cost of the recovered fissioning material after the first and subsequent reprocessing cycles taking account of the handling of radioactive wastes; Atomic Energy, Vol. 89, No. 4, 2000

Observation of the New Heavy Neutron-Rich Isotope 238Th

A new nuclide 238Th was produced via multinucleon transfer reaction by 60 MeV/u 18O ion irradiation of the natural uranium. The thorium was radiochemically separated from the mixture of uranium and reaction products. The growth and decay of 238Pa gamma rays was observed, identifying the 238Pa parent as 238Th. The half-life of 238Th was determined to be 9.4±2.0 min.

Towards in-situ ultrasonic transmission measurement of melt crusts above 2000 °C

Abstract Experiments are underway to assess the behaviour of mixtures of uranium and zirconium oxides and of steel, known as corium, in the 1500–3000 °C range. In the first section, a bibliographic review is presented. Speed of sound and acoustic attenuation of the various constituents of corium experiments has been experimentally determined up to 1780 °C. Large damping has been observed on pure UO 2 above 1100 °C. The onset of this dramatic attenuation is reported to be delayed to higher temperatures in the case of mixtures. An experimental apparatus is proposed to monitor the acoustic properties of the crust on the top of a crucible. It consists of two refractory metal waveguides which will be brought in contact with the crust surface at 5 cm intervals. Transmission between transducers on both waveguides will provide data on the crust acoustic behaviour. As a validation experiment, two waveguides have been immersed in molten lead and the transmission has been studied during the solidification. As with liquid lead, acoustic wetting has been observed in liquid mercury, enabling acoustic transmission between the two waveguides. This surprising effect has been observed over a large frequency range from 50 kHz to a few MHz.

Metallurgical Examination of Bore Samples from the Three Mile Island Unit 2 Reactor Core

Samples of the bores obtained from the melted core of the Three Mile Island Unit 2 (TMI-2) reactor were investigated as part of the TMI-2 accident evaluation program. The samples included fuel rod segments, control rod cladding, melted core rocks, and powder debris from various bores into the reactor core.The microstructure of the specimens was determined by examination and analysis of surface and cross section by means of electro-optical techniques (including energy-dispersive X-ray scanning electron microscopy and microprobe analysis). Gamma spectroscopy and a fission product release study were also performed on the specimens.The melted core rock specimens obtained from the G12 bores were mostly porous ceramic mixtures of uranium and zirconium oxides in the form of a fine eutectic of UO2- and ZrO2-rich phases and oxidized ferrous material derived from the stainless steel components. The rock specimens showed variations in porosity and ferrous content but a similar UO2-ZrO2 eutectic structure.Fission product analysis of the fuel segments and rocks indicated relatively low levels of activity due to fission products such as I37Cs, 106Ru, I54Eu, and fuel irradiation products. No volatile 129I was detected, but most other products displayed some retention in the melted core samples.According to whether eutectics were formed between Zr(O) and UO2 or ZrO2 and UO2 (i.e., the oxygen potential in the core), reference to the phase diagrams suggests temperatures of 2173 to 2873 K (1900 to 2600°C) for substantial periods and even up to 3073 K (2800°C) (U02 melting point) for the completely melted core specimens. Agglomerate specimens indicate lower temperatures [∼1673 K (1400°C), the melting point of stainless steel] and shorter excursion times at the edge of the melt zone. The remaining fuel rod segments showed very little change, indicating that the severe overheating is localized to the central part of the reactor pile.

Some reactions of uranium chloride pentafluoride

The molecule UF5Cl has been isolated, together with an excess of UF6, in a solid matrix of Ar, N2, or CO and characterised by its i.r. spectrum. Under these conditions it dissociates under the action of radiation having wavelengths close to 500 nm to give UF5; OCCl˙ and OCClF are also formed on photolysis in a solid CO matrix, whereas a species believed to be U2F11 is formed on photolysis in a solid N2 matrix. CCl3F solutions of fluoride-rich mixtures of uranium(VI) chloride fluorides have been shown to function as chlorinating, fluorinating, or chlorofluorinating reagents in their reactions with various unsaturated molecules at temperatures low enough to preclude thermal decomposition of the mixed halides (< –60 °C).

Monochromatic source of Lyman-α radiation

A source has been developed which produces a pure spectrum of Lyman-α radiation (1215.7 Å). This source incorporates a wavelength selective filter and an rf-excited helium-filled lamp containing a mixture of uranium and uranium hydride in a sidearm. The uranium serves as a getter to eliminate atmospheric contaminants, and the uranium hydride, when heated, supplies H2 in a reproducible manner. The filter consists of a flowing-oxygen cell and a narrowband interference filter. The distinctive advantage of this device is that radiation in the VUV at a well-defined wavelength is obtained without the use of a monochromator. Characteristics of the source and measurements of the irradiance of the spectral line are given for a typical lamp. The irradiance and spectral purity are seen to be not strongly dependent on oxygen flow.

Analysis of uranium and plutonium in binary mixtures using the technique of track registration from solutions

A method based on the registration of alpha-particle and fission-fragment tracks from solution media on cellulose nitrate (CN) and Lexan ∗ polycarbonate (PC) plastics, respectively, has been developed and tested for the determination of low concentrations of 235U and 239PU in a mixture The analysis is done by setting up two simultaneous equations relating the concentrations to the alpha and fission track densities obtained from the solutions of standard uranium, plutonium and the mixture. Results of analysis of a number of binary mixtures of uranium and plutonium are presented. The concentrations of 235U and 239Pu were in the range of a few micrograms per millilitre. The errors in the determinations were found to be below 5%. The possible application of the method for the analysis of uranium and plutonium in low-concentration process-waste solutions, including highly radioactive fuel-reprocessing waste streams, is discussed.

Volumetric determination of uranium in a mixture of uranium and plutonium employing Fe(III) as titrant

A method for the determination of U in the presence of Pu based on the reduction of U to U(IV) and Pu to Pu(III) by zinc amalgam followed by oxidimetry of U(IV) has been developed. Fe(III) perchlorate was chosen as the most suitable titrant for the selective oxidation of U(IV) as conventional oxidising titrants fail in the presence of Pu(III). The potentiometric titration of U(IV) with Fe(III) is known to be slow. This problem, however, has been overcome by selecting a suitable buffer medium and complexing agent to alter the potentials of the Fe(III)/Fe(II) and U(VI)/U(IV) systems in the favourable direction. Results of the titration of U(IV) with Fe(III) at pH 2 in the presence of ferrozine as complexing agent for Fe(II) are summarized. R.S.D. obtained for twenty determinations of 3–5 mg of U was 0.3 % in the presence of 1–4 mg of Pu.

History of Fissile Array Measurements in the United States

The first measurements with arrays of fissile metal were performed at the Los Alamos Scientific Laboratory in 1947 and the first with fissile solutions were performed at the Oak Ridge Critical Experiments Facility in 1949. Since then, there have been many other significant experiments at several U.S. laboratories including, in addition, Rocky Flats, Battelle-Pacific Northwest Laboratory, and the Lawrence Livermore Laboratory. Array tests were the primary sources of data used in developing criticality criteria for fissile-process plants, and they provided the basis for several empirical storage models that are still in use. Some of the experimental data also serve to validate Monte Carlo neutron transport calculations now used extensively by the nuclear safety engineer in the design of storage and processing facilities. It is felt that there should be additional experimental data for further validation of calculational methods relied upon for criticality safety evaluation. The deficient areas include low-/sup 235/U-enriched uranium, /sup 233/U with and without thorium, and plutonium--uranium mixtures. Also, critical data are lacking for arrays with the concrete reflectors normally found in process-plant environments, and additional experiments on concrete-reflected arrays are needed.

Differential spectrophotometric determination of micro amounts of metal ions in mixtures

A differential spectrophotometric method is described for the simultaneous determinations of iron and molybdenum, iron and titanium, iron and uranium, molybdenum and titanium, and molybdenum and uranium in mixtures, in a single solution at room temperature. The general method consists in treating the mixture with 5 ml of 1 M sodium acetate and 1 ml of 1% pyrocatechol in 25-ml volumetric flasks. The optical density of color of mixed complexes is determined at the corresponding pair of wavelengths suggested. From the optical density and molar extinction coefficient, the concentration of each ion in mixtures has been calculated. By this procedure, the metal ions can be determined even if they differ by large ratios. Procedures have been extended for the analysis of mixtures containing iron, molybdenum, and titanium (or uranium). These procedures can be applied for the analysis of any type of materials, like minerals, rocks, ores, alloys, steels and refractory materials. Spectra of each complex and their Beer's law were studied in detail. Experiments were also made to show that the calculated and observed optical density of mixture complexes is the same, thereby the additive of optical density of mixtures at each pair of wavelengths suggested for individual determinations of ions have been verified.

Uranium use in CANDU reactors—review and prospect

A co-operative governmentindustry development program launched in the late 1950s has generated in Canadian industry the capability to produce high-quality uranium oxide fuel bundles at all-inclusive prices of less than $50 per kg U. The direct consequence is that fuel cycle costs in heavy-water-moderated CANDU power reactors are the lowest of any reactor type now in operation. Even in competition with fast breeder reactors the CANDU systems are likely to retain the lowest fuelling costs for years to come.The Canadian requirements for sintered UO2 pellets will grow by nearly tenfold to some 1000 tons per year during the 1970s. Within that period, however, several alternative fuel cycles may offer stiff economic competition for natural UO2 once-through fuel. These include (a) the ‘valubreeder’ cycle which is based on a mixture of uranium and thorium; (b) natural uranium silicide, U3Si; and (c) plutonium recycle. All three alternatives become more attractive relative to natural UO2 as uranium prices rise. Résumé Un programme de développement organisé conjointement par le gouvernement et l'industrie dans les dernières années 1950 a donné à l'industrie canadienne la compétence voulue pour fabriquer des faisceaux combustibles de haute qualité à un prix global inférieur à $50 par kilogramme d'uranium. La conséquence directe est que le coût du cycle de combustible des réacteurs CANDU modérés par eau lourde est inférieur à celui de tous les types de réacteurs actuellement en service. Même en concurrence avec les réacteurs surgénérateurs la ftlière CANDU continuera probablement d'avoir pendant des années le cycle de combustible le plus économique.Au cours de la prochaine décennie, les besoins du Canada en matière de pastilles d'UO2 augmenteront d'environ 1000 tonnes par an. Au cours de cette période, cependant, plusieurs autres cycles de combustible pourraient concurrencer le combustible à passe unique qu'est le bioxyde d'uranium naturel. Il y aurait, entre autres: a) le cycle quasi-surgénérateur employant un mélange d'uranium et de thorium; b) le siliciure d'uranium naturel, U3Si; et c) le plutonium recyclé. Ces trois possibilités gagneront de l'intérêt par rapport au bioxyde d'uranium naturel à mesure que le prix de l'uranium augmentera.

Fluorometic Determination of Microquantities of Uranium in Mixtures of Uranium and Plutonium

Fluorometic Determination of Microquantities of Uranium in Mixtures of Uranium and Plutonium

Effect of Oxygen on the Vapor Pressure of Uranium

The published observations pertaining to the vapor pressure of uranium have been examined for consistencies. Since it was shown previously by Rauh and Thorn that oxygen suppresses the vapor pressure, the experiment by DeMaria, Burns, Drowart, and Inghram involving a mixture of uranium and aluminum oxide contained in a molybdenum effusion cell cannot yield the vapor pressure of pure uranium, but rather yields the individual partial pressures for the equilibrium U(Mo and O in solution)+UO2(s)=2UO(g),and a composition of the vapor which is approximately that of the monoxide. Using all the data one can construct a diagram schematically describing the effect of oxygen on the vapor pressure of uranium.

Study of Mechanism of Reaction Hot Pressing

The mechanism of reaction sintering was studied in mixtures of uranium‐uranium sesquinitride and uranium‐silicon. Refractory uranium mono‐nitride of high density may be prepared by hot pressing mixtures of uranium and uranium sesquinitride at 1000°C. Steps of the process are (1) densification of the uranium metal into a continuous uranium matrix and (2) reaction of the uranium sesquinitride with the uranium matrix. On the contrary, during hot pressing uranium‐silicon mixtures react before any densification of the uranium occurs. The pressing temperature required for a given density is then practically the same as for unreacted powders and the previously formed compound. Densification stages of the prereacted compounds may be described by the equations y = aot2/5+ b0, y = a1 log t + b1, and y = a2t−1/2+ b2, where y is the shrinkage and t is the time. The first stage is absent during densification of the powder mixtures.

Volumetric Determination of Uranium in the Presence of Niobium Utilizing a Bismuth Metal Reductor

The use of bismuth metal as a reducing agent during the determination of uranium in the presence of large amounts of niobium was studied. Other metals or metal amalgams reduce both uranium and niobium, prohibiting a simple volumetric deteimination. When a strongly acid solution bismuth metal, the uranium(VI) is quantitatively reduced formed and without encountering the problem of niobium hydrolysis. The uranium is determined indirectly by reaction with excess iron(III), producing iron(II) which is titrated with standard cerium(IV) solution. The end point with extraction of the cobalt--2-nitroso-1-naphthol complex into chloroform results in an accurate and reliable method of the titration is determined potentiometrically. Interferences were encountered with iron, molybdenum, vanadium, and tungsten in mixtures of uranium and niobium but not with nickel, zirconium, chromium, beryllium, copper, titanium, and aluminum.

THE SINTERING OF URANIUM-CARBON-IRON ALLOYS IN THE PRESENCE OF A LIQUID PHASE

The preparation of uranium mono carbide bodies containing ∼1% open porosity and with a density of ∼12·4 g/c.c. by cold compacting and sintering uranium/graphite mixtures is described. The effect of iron additions in producing a liquid phase during sintering has been studied, and it is shown that the addition of 10 wt.-% UFe2 to the stoichiometric mixture of uranium and graphite raises the sintered density to ∼13·1 g/c.c. and reduces the open porosity to ∼0·5%. The mechanism of carbide formation and densification in the presence of a liquid phase is discussed.