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Sample records for americium plutonium uranium

  1. Imitators of plutonium and americium in a mixed uranium- plutonium nitride fuel

    Science.gov (United States)

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

    2016-04-01

    Uranium nitride and mix uranium nitride (U-Pu)N is most popular nuclear fuel for Russian Fast Breeder Reactor. The works in hot cells associated with the radiation exposure of personnel and methodological difficulties. To know the main physical-chemical properties of uranium-plutonium nitride it necessary research to hot cells. In this paper, based on an assessment of physicochemical and thermodynamic properties of selected simulators Pu and Am. Analogues of Pu is are Ce and Y, and analogues Am - Dy. The technique of obtaining a model nitride fuel based on lanthanides nitrides and UN. Hydrogenation-dehydrogenation- nitration method of derived powders nitrides uranium, cerium, yttrium and dysprosium, held their mixing, pressing and sintering, the samples obtained model nitride fuel with plutonium and americium imitation. According to the results of structural studies have shown that all the samples are solid solution nitrides rare earth (REE) elements in UN.

  2. Distribution of uranium, americium and plutonium in the biomass of freshwater macrophytes

    Energy Technology Data Exchange (ETDEWEB)

    Zotina, T.A.; Kalacheva, G.S.; Bolsunovsky, A.YA. [Institute of Biophysics SB RAS, Akademgorodok, Krasnoyarsk (Russian Federation)

    2010-07-01

    Accumulation of uranium ({sup 238}U), americium ({sup 241}Am) and plutonium ({sup 242}Pu) and their distribution in cell compartments and biochemical components of the biomass of aquatic plants Elodea canadensis, Ceratophyllum demersum, Myrioplyllum spicatum and aquatic moss Fontinalis antipyretica have been investigated in laboratory batch experiments. Isotopes of uranium, americium and plutonium taken up from the water by Elodea canadensis apical shoots were mainly absorbed by cell walls, plasmalemma and organelles. A small portion of isotopes (about 6-13 %) could be dissolved in cytoplasm. The major portion (76-92 %) of americium was bound to cell wall cellulose-like polysaccharides of Elodea canadensis, Myriophyllum spicatum, Ceratophyllum demersum and Fontinalis antipyretica, 8-23 % of americium activity was registered in the fraction of proteins and carbohydrates, and just a small portion (< 1%) in lipid fraction. The distribution of plutonium in the biomass fraction of Elodea was similar to that of americium. Hence, americium and plutonium had the highest affinity to cellulose-like polysaccharides in Elodea biomass. Distribution of uranium in the biomass of Elodea differed essentially from that of transuranium elements: a considerable portion of uranium was recorded in the fraction of protein and carbohydrates (51 %). From our data we can assume that uranium has higher affinity to carbohydrates than proteins. (authors)

  3. Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

    International Nuclear Information System (INIS)

    Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O2 lattice in an irradiated (60 MW d kg−1) MOX sample was performed employing micro-X-ray fluorescence (µ-XRF) and micro-X-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (∼0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am3+ species within an [AmO8]13− coordination environment (e.g. >90%) and no (III XAFS spectra recorded for the irradiated MOX sub-sample in the rim zone for a 300 μm×300 μm beam size area investigated over six scans of 4 h. The records remain constant during multi-scan. The analysis of the XAFS signal shows that Am is found as trivalent in the UO2 matrix. This analytical work shall open the door of very challenging analysis (speciation of fission product and actinides) in irradiated nuclear fuels. - Highlights: • Americium was characterized by microX-ray absorption spectroscopy in irradiated MOX fuel. • The americium redox state as determined from XAS data of irradiated fuel material was Am(III). • In the sample, the Am3+ face an AmO813− coordination environment in the (Pu,U)O2 matrix. • The americium dioxide is reduced by the uranium dioxide matrix

  4. Characterization of uranium, plutonium, neptunium, and americium in HLW supernate for LLW certification

    International Nuclear Information System (INIS)

    The 1S Manual requires that High Level Waste (HLW) implement a waste certification program prior to sending waste packages to the E-Area vaults. To support the waste certification plan, the HLW supernate inventory of uranium, plutonium, neptunium and americium have been characterized. This characterization is based on the chemical, isotopic and radiological properties of these elements in HLW supernate. This report uses process knowledge, solubility data, isotopic inventory data and sample data to determine if any isotopes of the aforementioned elements will exceed the minimum reportable quantity (MRQ) for waste packages contaminated with HLW supernate. If the MRQ can be exceeded for a particular nuclide, then a method for estimating the waste package content is provided. Waste packages contaminated from HLW supernate do not contain sufficient U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242 or Am-241 to warrant separate reporting on the shipping manifest. Calculations show that, on average, more than 100 gallons of supernate is required to exceed the PAC (package acceptance criteria) for each of these nuclides. Thus it is highly unlikely that the PAC would be exceeded for these nuclides and unlikely that the MRQ would be exceeded. These nuclides should be manifested as zero for waste packages contaminated with HLW supernate. The only actinide isotopes that may exceed the MRQ are Np-237 and Pu-238. The recommended method to calculate the amount of these two isotopes in waste packages contaminated with HLW supernate is to ratio them to the measured Cs-137 activity

  5. Distribution, retention and dosimetry of plutonium and americium in the rat, dog and monkey after inhalation of an industrial-mixed uranium and plutonium oxide aerosol

    International Nuclear Information System (INIS)

    This study provides information on patterns of radiation dose in laboratory animals after inhalation exposure to an aerosol of one form of mixed uranium and plutonium oxide. The aerosol contained a mixture of UO2 and 750 deg C heat-treated PuO2 obtained from the ball milling operation in a mixed-oxide fuel fabrication process. Americium-241 from the decay of 241Pu was also present in the PuO2 matrix. Fischer-344 rats, Beagle dogs, and Cynomolgus and Rhesus monkeys inhaled aerosols re-generated from dry mixed oxide powders with particle size distribution characteristics similar to those observed in samples collected at the industrial site. Clearance from the lung and distribution in other tissues of the plutonium from this UO2 + PuO2 admixture was similar to what has been observed for PuO2 from laboratory-produced aerosols. The UO2-PuO2 aerosol was relatively insoluble in the lungs of all species. Monkeys and rats cleared plutonium and americium from their lungs faster than dogs. Very little plutonium or americium translocated within the first 2 yr after exposure to tissues other than tracheobronchial lymph nodes. The greater accumulation of plutonium and americium in the tracheobronchial lymph nodes of dogs as compared to monkeys and rats combined with the more rapid initial clearance of these radionuclides from the lungs of rats and monkeys suggests that errors could result from using data from a single animal species to estimate risk to humans from inhalation of these industrial aerosols. (author)

  6. Use of radioanalytical methods for determination of uranium, neptunium, plutonium, americium and curium isotopes in radioactive wastes

    International Nuclear Information System (INIS)

    Activated charcoal is a common type of radioactive waste that contains high concentrations of fission and activation products. The management of this waste includes its characterization aiming the determination and quantification of the specific radionuclides including those known as Difficult-to-Measure Radionuclides (RDM). The analysis of the RDM's generally involves complex radiochemical analysis for purification and separation of the radionuclides, which are expensive and time-consuming. The objective of this work was to define a methodology for sequential analysis of the isotopes of uranium, neptunium, plutonium, americium and curium present in a type of radioactive waste, evaluating chemical yield, analysis of time spent, amount of secondary waste generated and cost. Three methodologies were compared and validated that employ ion exchange (TI + EC), extraction chromatography (EC) and extraction with polymers (ECP). The waste chosen was the activated charcoal from the purification system of primary circuit water cooling the reactor IEA-R1. The charcoal samples were dissolved by acid digestion followed by purification and separation of isotopes with ion exchange resins, extraction and chromatographic extraction polymers. Isotopes were analyzed on an alpha spectrometer, equipped with surface barrier detectors. The chemical yields were satisfactory for the methods TI + EC and EC. ECP method was comparable with those methods only for uranium. Statistical analysis as well the analysis of time spent, amount of secondary waste generated and cost revealed that EC method is the most effective for identifying and quantifying U, Np, Pu, Am and Cm present in charcoal. (author)

  7. Use of radioactive methods for determination of uranium, neptunium, plutonium, americium and curium isotopes in waste radioactive

    International Nuclear Information System (INIS)

    Activated charcoal is a common type of radioactive waste that contains high concentrations of fission and activation products. The management of this waste includes its characterization aiming the determination and quantification of the specific radionuclides including those known as Difficult-to-Measure Radionuclides (RDM). The analysis of the RDM's generally involves complex radiochemical analysis for purification and separation of the radionuclides, which are expensive and time-consuming. The objective of this work was to define a methodology for sequential analysis of the isotopes of uranium, neptunium, plutonium, americium and curium present in a type of radioactive waste, evaluating chemical yield, analysis of time spent, amount of secondary waste generated and cost. Three methodologies were compared and validated that employ ion exchange (TI+EC), extraction chromatography (EC) and extraction with polymers (ECP). The waste chosen was the activated charcoal from the purification system of primary circuit water cooling the reactor IEA-R1. The charcoal samples were dissolved by acid digestion followed by purification and separation of isotopes with ion exchange resins, extraction and chromatographic extraction polymers. Isotopes were analyzed on an alpha spectrometer, equipped with surface barrier detectors. The chemical yields were satisfactory for the methods TI+EC and EC. ECP method was comparable with those methods only for uranium. Statistical analysis as well the analysis of time spent, amount of secondary waste generated and cost revealed that EC method is the most effective for identifying and quantifying U, Np, Pu, Am and Cm present in charcoal. (author)

  8. Inert matrices, uranium-free plutonium fuels and americium targets. Synthesis of CAPRA, SPIN and EFTTRA studies

    International Nuclear Information System (INIS)

    A first selection of inert-matrix materials, actinide support alone (Pu and Am based), and compound materials, U free plutonium burning fuels and heterogeneous americium targets are discussed. Basic properties, fabrication, and reprocessing studies, European in-pile and out-of-pile tests, performed recently in the framework of CAPRA, SPIN and EFTTRA programs, are reviewed here. Taking into account these studies and on the bases of the different requirements to be met in each of the fuels and targets, a number of materials have been selected as 'promising candidates'. Trends for further research on these materials are established. (author)

  9. Pyrochemical technology of plutonium and americium preparation and purification

    International Nuclear Information System (INIS)

    Pyrochemical tecnology of metallic plutonium and americium preparation and purification is considered. Investigations into plutonium dioxide reduction up to metal; plutonium electrolytic refining in molten salts; plutonium extraction from the molten salts and preparation of americium dioxide and metallic americium from its tetrafluoride are described

  10. Decontaminaion of metals containing plutonium and americium

    International Nuclear Information System (INIS)

    Melt-slagging (melt-refining) techniques were evaluated as a decontamination and consolidation step for metals contaminated with oxides of plutonium and americium. Experiments were performed in which mild steel, stainless steel, and nickel contaminated with oxides of plutonium and americium were melted in the presence of silicate slags of various compositions. The metal products were low in contamination, with the plutonium and americium strongly fractionated to the slags. Partition coefficients (plutonium in slag/plutonium in steel) of 7 x 106 were measured with boro-silicate slag and of 3 x 106 with calcium, magnesium silicate slag. Decontamination of metals containing as much as 14,000 ppM plutonium appears to be as efficient as for metals with plutonium levels of 400 ppM. Staged extraction, that is, a remelting of processed metal with clean slag, results in further decontamination of the metal. The second extraction is effective with either resistance-furnace melting or electric-arc melting. Slag adhering to the metal ingots and in defects within the ingots is in the important contributors to plutonium retained in processed metals. If these sources of plutonium are controlled, the melt-refining process can be used on a large scale to convert highly contaminated metals to homogeneous and compact forms with very low concentrations of plutonium and americium. A conceptual design of a melt-refining process to decontaminate plutonium- and americium-contaminated metals is described. The process includes single-stage refining of contaminated metals to produce a metal product which would have less than 10 nCi/g of TRU-element contamination. Two plant sizes were considered. The smaller conceptual plant processes 77 kg of metal per 8-h period and may be portable.The larger one processes 140 kg of metal per 8-h period, is stationary, and may be near te maximum size that is practical for a metal decontamination process

  11. Aqueous Chloride Operations Overview: Plutonium and Americium Purification/Recovery

    Energy Technology Data Exchange (ETDEWEB)

    Kimball, David Bryan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Skidmore, Bradley Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-06-22

    Acqueous Chloride mission is to recover plutonium and americium from pyrochemical residues (undesirable form for utilization and storage) and generate plutonium oxide and americium oxide. Plutonium oxide is recycled into Pu metal production flowsheet. It is suitable for storage. Americium oxide is a valuable product, sold through the DOE-OS isotope sales program.

  12. Actinide Oxidation State and O/M Ratio in Hypostoichiometric Uranium-Plutonium-Americium U0.750Pu0.246Am0.004O2-x Mixed Oxides.

    Science.gov (United States)

    Vauchy, Romain; Belin, Renaud C; Robisson, Anne-Charlotte; Lebreton, Florent; Aufore, Laurence; Scheinost, Andreas C; Martin, Philippe M

    2016-03-01

    Innovative americium-bearing uranium-plutonium mixed oxides U1-yPuyO2-x are envisioned as nuclear fuel for sodium-cooled fast neutron reactors (SFRs). The oxygen-to-metal (O/M) ratio, directly related to the oxidation state of cations, affects many of the fuel properties. Thus, a thorough knowledge of its variation with the sintering conditions is essential. The aim of this work is to follow the oxidation state of uranium, plutonium, and americium, and so the O/M ratio, in U0.750Pu0.246Am0.004O2-x samples sintered for 4 h at 2023 K in various Ar + 5% H2 + z vpm H2O (z = ∼15, ∼90, and ∼200) gas mixtures. The O/M ratios were determined by gravimetry, XAS, and XRD and evidenced a partial oxidation of the samples at room temperature. Finally, by comparing XANES and EXAFS results to that of a previous study, we demonstrate that the presence of uranium does not influence the interactions between americium and plutonium and that the differences in the O/M ratio between the investigated conditions is controlled by the reduction of plutonium. We also discuss the role of the homogeneity of cation distribution, as determined by EPMA, on the mechanisms involved in the reduction process. PMID:26907589

  13. Americium, plutonium and uranium contamination and speciation in well waters, streams and atomic lakes in the Sarzhal region of the Semipalatinsk Nuclear Test Site, Kazakhstan

    International Nuclear Information System (INIS)

    New data are reported on the concentrations, isotopic composition and speciation of americium, plutonium and uranium in surface and ground waters in the Sarzhal region of the Semipalatinsk Test Site, and an adjacent area including the settlement of Sarzhal. The data relate to filtered water and suspended particulate from (a) streams originating in the Degelen Mountains, (b) the Tel'kem 1 and Tel'kem 2 atomic craters, and (c) wells on farms located within the study area and at Sarzhal. The measurements show that 241Am, 239,240Pu and 238U concentrations in well waters within the study area are in the range 0.04-87 mBq dm-3, 0.7-99 mBq dm-3, and 74-213 mBq dm-3, respectively, and for 241Am and 239,240Pu are elevated above the levels expected solely on the basis of global fallout. Concentrations in streams sourced in the Degelen Mountains are similar, while concentrations in the two water-filled atomic craters are somewhat higher. Suspended particulate concentrations in well waters vary considerably, though median values are very low, at 0.01 mBq dm-3, 0.08 mBq dm-3 and 0.32 mBq dm-3 for 241Am, 239,240Pu and 238U, respectively. The 235U/238U isotopic ratio in almost all well and stream waters is slightly elevated above the 'best estimate' value for natural uranium worldwide, suggesting that some of the uranium in these waters is of test-site provenance. Redox analysis shows that on average most of the plutonium present in the microfiltered fraction of these waters is in a chemically reduced form (mean 69%; 95% confidence interval 53-85%). In the case of the atomic craters, the proportion is even higher. As expected, all of the americium present appears to be in a reduced form. Calculations suggest that annual committed effective doses to individual adults arising from the daily ingestion of these well waters are in the range 11-42 μSv (mean 21 μSv). Presently, the ground water feeding these wells would not appear to be contaminated with radioactivity from past

  14. Recovery of americium-241 from raffinates of plutonium purification columns

    International Nuclear Information System (INIS)

    Recovery and purification of americium from ion exchange raffinates generated during purification of aged plutonium is described. The method consists of the following stages: (i) co-precipitation of americium with kilogramme quantities of rare earth oxalates, (ii) destruction of oxalate and removal of residual plutonium from nitric acid medium using anion exchange process, (iii) preliminary separation of americium making use of its preferential uptake on an anion exchange column from thiocyanate medium and (iv) extraction of americium and remaining rare earths into di-(2-ethyl hexyl) phosphoric acid followed by preferential back washing of americium by lactic acid medium containing DTPA. (author)

  15. Plutonium and americium in soil organic matter

    International Nuclear Information System (INIS)

    A gley soil from west Cumbria, with specific activities in its surface horizon of 5-10 kBq kg-1239,240Pu and comparable 241Am levels, has been used as a source of actinide-enriched organic fractions. Humic and fulvic acids were isolated by conventional alkali extraction and investigated by gel filtration, treatment with organic solvents and differential flocculation procedures. All these techniques are capable of resolving the organics into two or more fractions, with specific activities up to 80 kBq kg-239,240Pu. There is evidence for differentiation of plutonium and americium, with americium being concentrated, to some extent, in the lower molecular weight fractions from gel filtration. (author)

  16. Solubility of Plutonium (IV) Oxalate During Americium/Curium Pretreatment

    Energy Technology Data Exchange (ETDEWEB)

    Rudisill, T.S.

    1999-08-11

    Approximately 15,000 L of solution containing isotopes of americium and curium (Am/Cm) will undergo stabilization by vitrification at the Savannah River Site (SRS). Prior to vitrification, an in-tank pretreatment will be used to remove metal impurities from the solution using an oxalate precipitation process. Material balance calculations for this process, based on solubility data in pure nitric acid, predict approximately 80 percent of the plutonium in the solution will be lost to waste. Due to the uncertainty associated with the plutonium losses during processing, solubility experiments were performed to measure the recovery of plutonium during pretreatment and a subsequent precipitation process to prepare a slurry feed for a batch melter. A good estimate of the plutonium content of the glass is required for planning the shipment of the vitrified Am/Cm product to Oak Ridge National Laboratory (ORNL).The plutonium solubility in the oxalate precipitation supernate during pretreatment was 10 mg/mL at 35 degrees C. In two subsequent washes with a 0.25M oxalic acid/0.5M nitric acid solution, the solubility dropped to less than 5 mg/mL. During the precipitation and washing steps, lanthanide fission products in the solution were mostly insoluble. Uranium, and alkali, alkaline earth, and transition metal impurities were soluble as expected. An elemental material balance for plutonium showed that greater than 94 percent of the plutonium was recovered in the dissolved precipitate. The recovery of the lanthanide elements was generally 94 percent or higher except for the more soluble lanthanum. The recovery of soluble metal impurities from the precipitate slurry ranged from 15 to 22 percent. Theoretically, 16 percent of the soluble oxalates should have been present in the dissolved slurry based on the dilution effects and volumes of supernate and wash solutions removed. A trace level material balance showed greater than 97 percent recovery of americium-241 (from the beta dec

  17. Solubility of Plutonium (IV) Oxalate During Americium/Curium Pretreatment

    International Nuclear Information System (INIS)

    Approximately 15,000 L of solution containing isotopes of americium and curium (Am/Cm) will undergo stabilization by vitrification at the Savannah River Site (SRS). Prior to vitrification, an in-tank pretreatment will be used to remove metal impurities from the solution using an oxalate precipitation process. Material balance calculations for this process, based on solubility data in pure nitric acid, predict approximately 80 percent of the plutonium in the solution will be lost to waste. Due to the uncertainty associated with the plutonium losses during processing, solubility experiments were performed to measure the recovery of plutonium during pretreatment and a subsequent precipitation process to prepare a slurry feed for a batch melter. A good estimate of the plutonium content of the glass is required for planning the shipment of the vitrified Am/Cm product to Oak Ridge National Laboratory (ORNL).The plutonium solubility in the oxalate precipitation supernate during pretreatment was 10 mg/mL at 35 degrees C. In two subsequent washes with a 0.25M oxalic acid/0.5M nitric acid solution, the solubility dropped to less than 5 mg/mL. During the precipitation and washing steps, lanthanide fission products in the solution were mostly insoluble. Uranium, and alkali, alkaline earth, and transition metal impurities were soluble as expected. An elemental material balance for plutonium showed that greater than 94 percent of the plutonium was recovered in the dissolved precipitate. The recovery of the lanthanide elements was generally 94 percent or higher except for the more soluble lanthanum. The recovery of soluble metal impurities from the precipitate slurry ranged from 15 to 22 percent. Theoretically, 16 percent of the soluble oxalates should have been present in the dissolved slurry based on the dilution effects and volumes of supernate and wash solutions removed. A trace level material balance showed greater than 97 percent recovery of americium-241 (from the beta dec

  18. Pyrochemical investigations into recovering plutonium from americium extraction salt residues

    International Nuclear Information System (INIS)

    Progress into developing a pyrochemical technique for separating and recovering plutonium from spent americium extraction waste salts has concentrated on selective chemical reduction with lanthanum metal and calcium metal and on the solvent extraction of americium with calcium metal. Both techniques are effective for recovering plutonium from the waste salt, although neither appears suitable as a separation technique for recycling a plutonium stream back to mainline purification processes. 17 refs., 13 figs., 2 tabs

  19. Salvage of plutonium-and americium-contaminated metals

    International Nuclear Information System (INIS)

    Melt-slagging techniques were evaluated as a decontamination and consolidation step for metals contaminated with oxides of plutonium and americium. Experiments were performed in which mild steel, stainless steel, and nickel metals contaminated with oxides of plutonium and americium were melted in the presence of silicate slags of various compositions. The metal products were low in contamination, with the plutonium and americium strongly fractionated to the slags. Partition coefficients (plutonium in slag/plutonium in steel) of 7*10/sup 6/ with borosilicate slag and 3*10/sup 6/ for calcium, magnesium silicate slag were measured. Decontamination of metals containing as much as 14,000 p.p.m. plutonium appears to be as efficient as that of metals with plutonium levels of 400 p.p.m. Staged extraction, that is, a remelting of processed metal with clean slag, results in further decontamination of the metal. 10 refs

  20. Update of JAEA-TDB. Additional selection of thermodynamic data for solid and gaseous phases on nickel, selenium, zirconium, technetium, thorium, uranium, neptunium plutonium and americium, update of thermodynamic data on iodine, and some modifications

    International Nuclear Information System (INIS)

    We additionally selected thermodynamic data for solid and gaseous phases of nickel, selenium, zirconium, technetium, thorium, uranium, neptunium, plutonium and americium to our thermodynamic database JAEA-TDB for geological disposal of radioactive waste of high-level and TRU wastes. We thermodynamically obtained equilibrium constant from addition and subtraction of Gibbs free energy of formation on nickel, selenium, zirconium, technetium, thorium, uranium, neptunium plutonium and americium, which were selected in the Thermochemical Database Project by the Nuclear Energy Agency in the Organisation for Economic Co-operation and Development. Furthermore, we collected and updated thermodynamic data on iodine, changed master species of technetium(IV), and added thermodynamic data on selenium due to improving reliability of the thermodynamic database. We prepared text files of the updated thermodynamic database (JAEA-TDB) for geochemical calculation programs of PHREEQC, EQ3/6 and Geochemist's Workbench. These text files are contained in the attached CD-ROM and will be available on our Website (http://migrationdb.jaea.go.jp/). (author)

  1. Mixed chelation therapy for removal of plutonium and americium

    International Nuclear Information System (INIS)

    Iron-binding compounds, 2,3-dihydroxybenzoic acid (DHBA), 2-hydroxybenzoic acid (HBA), and 2-(acetyloxy)benzoic acid (ABA), were tested for their ability to remove americium and plutonium from rats following intraperitioneal injection of the radionuclides as citrates (pH 5). Treatments, 2 mmol/kg, were given on days 3, 6, 10, 12 and 14 following the actinide injection. DHBA and HBA caused about a 20% decrease in liver retention of americium compared to the control value, and DHB caused a similar effect for plutonium. The above agents, co-administered with 0.5 mmol polyaminopolycarboxylic acid (PAPCA)-type chelons, did not change tissue retention of americium and plutonium from that due to the PAPCAs alone. Administration of americium and plutonium to the same rats is useful for studying removal agents since the two actinides behave independently in their biological disposition and response to removal

  2. Plutonium and Americium Geochemistry at Hanford: A Site Wide Review

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Felmy, Andrew R.

    2012-08-23

    This report was produced to provide a systematic review of the state-of-knowledge of plutonium and americium geochemistry at the Hanford Site. The report integrates existing knowledge of the subsurface migration behavior of plutonium and americium at the Hanford Site with available information in the scientific literature regarding the geochemistry of plutonium and americium in systems that are environmentally relevant to the Hanford Site. As a part of the report, key research needs are identified and prioritized, with the ultimate goal of developing a science-based capability to quantitatively assess risk at sites contaminated with plutonium and americium at the Hanford Site and the impact of remediation technologies and closure strategies.

  3. Determination of plutonium, uranium and americium/curium isotopes in environmental samples with anion exchange, UTEVA, Sr and DGA resin

    International Nuclear Information System (INIS)

    This study presents a quantitative sequential radiochemical separation method for the Pu, U, Sr and Am/Cm isotopes with an anion exchange resin, UTEVA resin, Sr resin and DGA resin in environmental samples. After the radionuclides were leached from samples with 8M HNO3, the Pu, U, Sr and Am/Cm isotopes were sequentially adsorbed on the anion exchange column, UTEVA column connected with Sr Spec column and DGA column. The Pu isotopes were purified from other nuclides through the anion exchange column, and the uranium isotopes were separated from other nuclides through the UTEVA column. Also, 90Sr was separated from other hindrance elements such as Ca2+, Ba2+ and Y3+ with the Sr Spec column. Finally, Am/Cm fractions were purified with the DGA and anion exchange resins. After α source preparation for the purified Pu, U and Am/Cm isotopes with the micro-coprecipitation method, the Pu, U and Am/Cm isotopes were measured by an alpha spectrometry. Strontium-90 was measured by a low level liquid scintillation counter. The radiochemical procedure for Pu, U and Am nuclides investigated in this study has been validated by application to IAEA Reference soils. (orig.)

  4. Investigations of neutron characteristics for salt blanket models; integral fission cross section measurements of neptunium, plutonium, americium and curium isotopes

    International Nuclear Information System (INIS)

    Neutron characteristics of salt blanket micromodels containing eutectic mixtures of sodium, zirconium, and uranium fluorides were measured on FKBN-2M, BIGR and MAKET facilities. The effective fission cross sections of neptunium, plutonium, americium, and curium isotopes were measured on the neutron spectra formed by micromodels. (author)

  5. Use of radioactive methods for determination of uranium, neptunium, plutonium, americium and curium isotopes in waste radioactive; Utilizacao de metodos radioanaliticos para determinacao de isotopos de uranio, netunio, plutonio, americio e curio em rejeitos radioativos

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Bianca

    2012-07-01

    Activated charcoal is a common type of radioactive waste that contains high concentrations of fission and activation products. The management of this waste includes its characterization aiming the determination and quantification of the specific radionuclides including those known as Difficult-to-Measure Radionuclides (RDM). The analysis of the RDM's generally involves complex radiochemical analysis for purification and separation of the radionuclides, which are expensive and time-consuming. The objective of this work was to define a methodology for sequential analysis of the isotopes of uranium, neptunium, plutonium, americium and curium present in a type of radioactive waste, evaluating chemical yield, analysis of time spent, amount of secondary waste generated and cost. Three methodologies were compared and validated that employ ion exchange (TI+EC), extraction chromatography (EC) and extraction with polymers (ECP). The waste chosen was the activated charcoal from the purification system of primary circuit water cooling the reactor IEA-R1. The charcoal samples were dissolved by acid digestion followed by purification and separation of isotopes with ion exchange resins, extraction and chromatographic extraction polymers. Isotopes were analyzed on an alpha spectrometer, equipped with surface barrier detectors. The chemical yields were satisfactory for the methods TI+EC and EC. ECP method was comparable with those methods only for uranium. Statistical analysis as well the analysis of time spent, amount of secondary waste generated and cost revealed that EC method is the most effective for identifying and quantifying U, Np, Pu, Am and Cm present in charcoal. (author)

  6. Gut uptake factors for plutonium, americium and curium

    International Nuclear Information System (INIS)

    Data on estimates of the absorption of plutonium, americium and curium from the human gut based on measurements of uptake in other mammalian species are reviewed. It is proposed that for all adult members of the public ingesting low concentrations of plutonium in food and water, 0.05% would be an appropriate value of absorption except when the conditions of exposure are known and a lower value can be justified. For dietary intakes of americium and curium, the available data do not warrant a change from the ICRP value of 0.05%. For newborn children ingesting americium, curium and soluble forms of plutonium, a value of 1% absorption is proposed for the first 3 months of life during which the infant is maintained on a milk diet. It is proposed that a value of 0.5% should be used for the first year of life to take account of the gradual maturation of the gut. In considering the ingestion of insoluble oxides of plutonium by infants, it is proposed that absorption is taken as 0.1% for the first 3 months and 0.05% for the first year. (author)

  7. Ingestion Pathway Transfer Factors for Plutonium and Americium

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, A.

    1999-07-28

    Overall transfer factors for major ingestion pathways are derived for plutonium and americium. These transfer factors relate the radionuclide concentration in a given foodstuff to deposition on the soil. Equations describing basic relationships consistent with Regulatory Guide 1.109 are followed. Updated values and coefficients from IAEA Technical Reports Series No. 364 are used when a available. Preference is given to using factors specific to the Savannah River Site.

  8. Liquid-liquid extraction separation and determination of plutonium and americium

    International Nuclear Information System (INIS)

    A procedure is described for the determination of plutonium and americium after their initial separation on barium sulfate. The barium sulfate is dissolved in perchloric acid and the antinides and lanthanides are extracted into bis(2-ethylhexyl)phosphoric acid (HDEHP). Americium along with other tervalent actinides and lanthanides is stripped from HDEHP with nitric acid. The lanthanides are removed on a column of HDEHP supported on Teflon powder, and the americium and other tervalent actinides are electrodeposited for their determination by α spectrometry. The plutonium is stripped with nitric acid after reduction to the tervalent state with 2,5-di-tert-butylhydroquinone and electrodeposited for α spectrometry. Decontamination factors for plutonium and americium from each other and from other α emitters are 104 to 105. Two hours are required for the liquid-liquid extraction separations of plutonium and americium from eight samples. Recoveries of americium and plutonium through the HDEHP separatons are 99% and 95%, respectively

  9. Plutonium and americium in sediments of Lithuanian lakes

    International Nuclear Information System (INIS)

    The assessment of contribution of the global and the Chernobyl NPP (Nuclear Power Plant) accident plutonium and americium to plutonium pollution in sediments of Lithuanian lakes is presented. Theoretical evaluation of activity ratios of 238Pu/239+240Pu and 241Pu/239+240Pu in the reactor of unit 4 of the Chernobyl NPP before the accident was performed by means of the ORIGEN-ARP code from the SCALE 4.4A program package. Non-uniform distribution of radionuclides in depositions on the Lithuanian territory after nuclear weapon tests and the Chernobyl NPP accident is experimentally observed by measuring the lake sediment pollution with actinides. The activity concentration of sediments polluted with plutonium ranges from 2.0 ± 0.5 Bq/kg d.w. (dry weight) in Lake Asavelis to 14 ± 2 Bq/kg d.w. in Lake Juodis. The ratio of activity concentrations of plutonium isotopes 238Pu/239+240Pu measured by α-spectrometry in the 10-cm-thick upper layer of bottom sediment varies from 0.03 in Lake Juodis to 0.3 in Lake Zuvintas. The analysis of the ratio values shows that the deposition of the Chernobyl origin plutonium is prevailing in southern and south-western regions of Lithuania. Plutonium of nuclear weapon tests origin in sediments of lakes is observed on the whole territory of Lithuania, and it is especially distinct in central Lithuania. The americium activity due to 241Pu decay after the Chernobyl NPP accident and global depositions in bottom sediments of Lithuanian lakes has been evaluated to be from 0.9 to 5.7 Bq/kg. (author)

  10. Airborne plutonium-239 and americium-241 concentrations measured from the 125-meter Hanford Meteorological Tower

    International Nuclear Information System (INIS)

    Airborne plutonium-239 and americium-241 concentrations and fluxes were measured at six heights from 1.9 to 122 m on the Hanford meteorological tower. The data show that plutonium-239 was transported on nonrespirable and small particles at all heights. Airborne americium-241 concentrations on small particles were maximum at the 91 m height

  11. Uranium plutonium oxide fuels

    International Nuclear Information System (INIS)

    Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

  12. Sequential isotopic determination of plutonium, thorium, americium and uranium in the air filter and drinking water samples around the WIPP site

    International Nuclear Information System (INIS)

    The simultaneous determination of actinides in air filter and water samples around the WIPP site have been demonstrated. The analytical method is based on the selective separation and purification by anion exchange and Eichrome-TEVA, TRU and DGA-resin followed by determination of actinides by alpha spectrometry. Counting sources for alpha spectrometric measurements were prepared by microcoprecipitation on neodymium fluoride (NdF3). Radiochemical yields were determined using 242Pu, 229Th, 243Am and 232U as tracers. The validation of the method is performed through the analysis of reference materials or participating in laboratory intercomparison programs. The plutonium concentrations in aerosols varied seasonally, being highest in spring and summer due to the spring-time enhanced wind-storm transportation of radioactive aerosols from the stratosphere to the troposphere. The 238Pu/239+240Pu activity ratio in the aerosol samples is typically close to that of global fallout from historic above-ground nuclear weapons testing. The results presented here indicate that the source of plutonium in the WIPP environment results mainly from global nuclear fallout and there is no evidence of increases in radiological contaminants in the region that could be attributed to releases from the WIPP. (author)

  13. Placental transfer of americium and plutonium in mice

    International Nuclear Information System (INIS)

    Actinide element release to the environment and subsequent transfer through food chains to pregnant women may present a radiation hazard to fetuses in utero. To measure americium incorporation, four groups of pregnant mice were intravenously dosed with four concentrations of 243Am citrate in late pregnancy. Concentrations of 243Am in fetuses, placentas, and maternal femur, liver, carcass and pelt were determined 48 hr after injection. Doses were chosen so that the number of atoms of 243Am in each injected dose was equal to the number of atoms of 239Pu used in an earlier study of transplacental movement. Results indicate that, atom for atom, americium is incorporated into fetal tissue in lesser amounts (10-25 times) than is plutonium when intravenously administered to pregnant mice in equal atom amounts. Tissue analyses indicated that, at low dose levels, the average fraction of the dose incorporated into the fetuses decreased as the dose to the pregnant mouse was increased. A similar pattern was noted for placentas and maternal femurs. Data indicate that one must make extrapolations from low dose data only to make reasonable and realistic estimates of the transplacental movement and fetal incorporation of environmental levels of actinide elements in man and other species. (author)

  14. Kinetic parameters of transformation of americium and plutonium physicochemical forms in podsol soils

    International Nuclear Information System (INIS)

    Kinetic parameters of transformation of americium and plutonium physicochemical forms have been estimated and the prognosis of fixing and remobilization of these nuclides in podsol soils have been made on that basis in the work. (authors)

  15. Calibration procedures for in vivo sodium iodide spectrometry of plutonium and americium in the human lung

    International Nuclear Information System (INIS)

    This paper describes the calibration techniques and associated error analysis for the in vivo measurement by NaI spectrometry of heavy elements in the lung, specifically plutonium and americium. A very brief description of the instrumentation system is included

  16. Bidentate organophosphorus extraction of americium and plutonium from Hanford Plutonium Reclamation Facility waste

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, W.W.

    1974-09-01

    Applicability of bidentate organiphosphorus reagents to recovery of americium and plutonium from Hanford's Plutonium Reclamation Facility acid (approx. 2M HNO/sub 3/) waste stream (CAW solution) was studied. A solvent extraction process which employs a 30% DHDECMP (dihexyl-N, N-diethylcarbamylmethylene phosphonate)-CCl/sub 4/ extractant was devised and successfully tested in mixer-settler runs with actual CAW solution. Substitution of DHDECMP for DBBP eliminates the need to perform careful neutralization of unbuffered CAW soluton and increases overall americium recovery from the present 60 to 80% level to greater than or equal to 90%. Disadvantages to such substitution include the high cost (approx. $50/liter) of DHDECMP and the need to purify it (by acid (6M HCl) hydrolysis and alkaline washing) from small amounts of an unidentified impurity which prevents stripping of americium with dilute HNO/sub 3/. Distribution data obtained in this study confirm Siddall's earlier contention that bidentate organophosphorus regents can be used to remove actinides from concentrated high-level Purex process acid waste; a conceptual flowsheet for such an extraction process is given.

  17. Selective leaching studies of deep-sea sediments loaded with americium, neptunium and plutonium

    International Nuclear Information System (INIS)

    A series of selective leaching experiments were undertaken to investigate the solid phase speciation and distribution of americium, neptunium and plutonium which had been experimentally loaded onto different marine sediment types. The chemical leaches employed showed rather poor selectivity but certain trends were evident. Adsorption was not by ion exchange. Americium showed a preferential affinity for carbonate and plutonium for organic matter. Neptunium appeared to have no preferential affinities. Americium was sorbed by acetic acid residues (CaCO3 removed) and by unleached carbonate-rich sediments with equal efficiency. This indicates that it is able to diversify its solid phase affinity/distribution depending upon which solid phases are available. (author)

  18. Relativistic density functional theory modeling of plutonium and americium higher oxide molecules

    Science.gov (United States)

    Zaitsevskii, Andréi; Mosyagin, Nikolai S.; Titov, Anatoly V.; Kiselev, Yuri M.

    2013-07-01

    The results of electronic structure modeling of plutonium and americium higher oxide molecules (actinide oxidation states VI through VIII) by two-component relativistic density functional theory are presented. Ground-state equilibrium molecular structures, main features of charge distributions, and energetics of AnO3, AnO4, An2On (An=Pu, Am), and PuAmOn, n = 6-8, are determined. In all cases, molecular geometries of americium and mixed plutonium-americium oxides are similar to those of the corresponding plutonium compounds, though chemical bonding in americium oxides is markedly weaker. Relatively high stability of the mixed heptoxide PuAmO7 is noticed; the Pu(VIII) and especially Am(VIII) oxides are expected to be unstable.

  19. Adsorption-desorption characteristics of plutonium and americium with sediment particles in the estuarine environment: studies using plutonium-237 and americium-241

    International Nuclear Information System (INIS)

    The particle formation of plutonium and americium, their adsorption onto fresh water sediments and the desorption from the sediments in sea water were studied in the Laboratory under simulated river-estuary conditions, using γ-emitting plutonium-237 and americium-241. The results of the experiments show that the particle formation of plutonium depends on its valence states, on pH and on the salinity of the medium. For river water at pH4, some 25%, 20% and 30% of the added 237Pu was in particulate form, larger than 0.45 μm, for Pu (III), Pu (IV) and Pu (VI), respectively, while 65%, 90% and 50% of the respective valence states was associated with particles at pH 8. In sea water the general pattern remains similar, although Pu (VI) is more soluble in sea water owing to higher ligand concentrations for carbonate and bicarbonate complexes. The pH-dependency of particle formation of Am (III) is more steep than that of plutonium and seems to be influenced by colloidal substances occurring in the experimental media. The adsorption-desorption characteristics of plutonium and americium with the sediment in river water as well as sea water reflect the characteristics of their particle formation, being dependent upon such properties as valence states, the pH and salinity of the medium. A sewage effluent added to the media has small but measurable effects on the adsorption-desorption processes of plutonium. (author)

  20. Development of analytical methods for the separation of plutonium, americium, curium and neptunium from environmental samples

    OpenAIRE

    Salminen, Susanna

    2009-01-01

    In this work, separation methods have been developed for the analysis of anthropogenic transuranium elements plutonium, americium, curium and neptunium from environmental samples contaminated by global nuclear weapons testing and the Chernobyl accident. The analytical methods utilized in this study are based on extraction chromatography. Highly varying atmospheric plutonium isotope concentrations and activity ratios were found at both Kurchatov (Kazakhstan), near the former Semipalatinsk...

  1. Plutonium in depleted uranium penetrators

    International Nuclear Information System (INIS)

    Depleted Uranium (DU) penetrators used in the recent Balkan conflicts have been found to be contaminated with trace amounts of transuranic materials such as plutonium. This contamination is usually a consequence of DU fabrication being carried out in facilities also using uranium recycled from spent military and civilian nuclear reactor fuel. Specific activities of 239+240 Plutonium generally in the range 1 to 12 Bq/kg have been found to be present in DU penetrators recovered from the attack sites of the 1999 NATO bombardment of Kosovo. A DU penetrator recovered from a May 1999 attack site at Bratoselce in southern Serbia and analysed by University College Dublin was found to contain 43.7 +/- 1.9 Bq/kg of 239+240 Plutonium. This analysis is described. An account is also given of the general population radiation dose implications arising from both the DU itself and from the presence of plutonium in the penetrators. According to current dosimetric models, in all scenarios considered likely ,the dose from the plutonium is estimated to be much smaller than that due to the uranium isotopes present in the penetrators. (author)

  2. Synthesis and characterization of uranium-americium mixed oxides

    International Nuclear Information System (INIS)

    Americium isotopes represent a significant part of high-level and long-lived nuclear waste in spent fuels. Among the envisaged reprocessing scenarios, their transmutation in fast neutron reactors using uranium-americium mixed-oxide pellets (U1-xAmxO2±δ) is a promising option which would help decrease the ecological footprint of ultimate waste repository sites. In this context, this thesis is dedicated to the study of such compounds over a wide range of americium contents (7.5 at.% ≤ Am/(U+Am) ≤ 70 at.%), with an emphasis on their fabrication from single-oxide precursors and the assessment of their structural and thermodynamic stabilities, also taking self-irradiation effects into account. Results highlight the main influence of americium reduction to Am(+III), not only on the mechanisms of solid-state formation of the U1-xAmxO2±δ solid solution, but also on the stabilization of oxidized uranium cations and the formation of defects in the oxygen sublattice such as vacancies and cub-octahedral clusters. In addition, the data acquired concerning the stability of U1-xAmxO2±δ compounds (existence of a miscibility gap, vaporization behavior) were compared to calculations based on new thermodynamic modelling of the U-Am-O ternary system. Finally, α-self-irradiation-induced structural effects on U1-xAmxO2±δ compounds were analyzed using XRD, XAS and TEM, allowing the influence of americium content on the structural swelling to be studied as well as the description of the evolution of radiation-induced structural defects. (author)

  3. Removal of americium from effluent generated during the purification of plutonium by anion exchange

    Energy Technology Data Exchange (ETDEWEB)

    Noronha, Donald M.; Pius, Illipparambil C.; Chaudhury, Satyajeet [Bhabha Atomic Research Centre, Mumbai (India). Fuel Chemistry Div.

    2015-07-01

    Studies have been carried out on removal of americium from the effluent generated during anion exchange purification of plutonium. Americium 241, generated by the beta decay of Plutonium-241, is the major source of a activity in this highly acidic effluent and its removal would render the waste easily disposable. A simple and effective co-precipitation method, using thorium oxalate has been investigated for the treatment of this alpha active aqueous waste. Experiments have been carried out to identify optimum conditions to obtain high percentage co-precipitation with minimum amount of co-precipitant. Efforts were carried out to correlate the optimum conditions of co-precipitation of americium obtained in these experiments with solubility of thorium oxalate and americium oxalate calculated from solubility products of these compounds, stability constants of thorium and americium oxalate complexes taken from literature. The saturation capacity of thorium oxalate for Am(III) was also calculated by analyzing the K{sub d} value data using Langmuir adsorption equation. The strong tendency of americium to get co-precipitated and the high capacity exhibited by thorium oxalate for the uptake of americium indicate feasibility of using this method for the treatment of anion exchange effluent.

  4. Removal of americium from effluent generated during the purification of plutonium by anion exchange

    International Nuclear Information System (INIS)

    Studies have been carried out on removal of americium from the effluent generated during anion exchange purification of plutonium. Americium 241, generated by the beta decay of Plutonium-241, is the major source of a activity in this highly acidic effluent and its removal would render the waste easily disposable. A simple and effective co-precipitation method, using thorium oxalate has been investigated for the treatment of this alpha active aqueous waste. Experiments have been carried out to identify optimum conditions to obtain high percentage co-precipitation with minimum amount of co-precipitant. Efforts were carried out to correlate the optimum conditions of co-precipitation of americium obtained in these experiments with solubility of thorium oxalate and americium oxalate calculated from solubility products of these compounds, stability constants of thorium and americium oxalate complexes taken from literature. The saturation capacity of thorium oxalate for Am(III) was also calculated by analyzing the Kd value data using Langmuir adsorption equation. The strong tendency of americium to get co-precipitated and the high capacity exhibited by thorium oxalate for the uptake of americium indicate feasibility of using this method for the treatment of anion exchange effluent.

  5. Micrometallurgy of plutonium and uranium

    International Nuclear Information System (INIS)

    The article prepared but not published in 1958 as the report on the 2. Geneva conference on peaceful use of atomic energy is presented. The article contains data on preparation pioneered in the USSR trace mounts of metal plutonium and uranium and is great scientific and historical interest. Procedures on the preparation of metal halide salts, investigations on the development of protective chambers, equipment for the preparation of salts, equipment for the reduction, refractory ceramics as well as investigation into the operating regime, feature required for the purity of metal are performed. The developed strategy for the preparation of trace amounts of uranium and plutonium salts and procedure for their reduction by alkaline earth metal vapors provide a useful preparation of other metal trace amounts

  6. About the reaction between uranium-americium mixed oxides and sodium

    International Nuclear Information System (INIS)

    The recycling and fission of the highly toxic minor actinides neptunium and americium is only possible in a liquid metal cooled fast breeder reactor, for nuclear physical reasons. The present work is part of a research program dealing with the fuel-coolant interaction. Fuel pellets with equal parts of americium and uranium and varying oxygen-metal ratio were investigated. A behaviour comparable to that of uranium-plutonium mixed oxides was suggested as a first approach. The reaction of sodium with (U0.5Am0.5)O2-x results in a complete desintegration of the sintered pellet whereas (U, Pu)O2-x pellets show a small increase in volume. A first explanation of the strong reaction of uranium-americium mixed oxides compared to (U, Pu)O2-x or (U, Np)O2-x could be provided by the less negative oxygen potential of the former. Ternary and polynary oxides which are possible products of the fuel-coolant reaction were prepared and characterised by X-ray diffraction. Their oxygen potentials were measured using a solid state e.m.f. cell. Neither Na2AmO3 nor Na3AmO4 can coexist with sodium metal. The measured ΔGO2 values of the Am(IV) and Am (V)-compounds are much higher than those of the sodium uranates(VI) or sodium neptunates(VI). Only Na2O seems to be likely as product of the fuel-coolant interactions. It could be determined in reacted samples by X-ray diffraction. The relatively high oxygen potentials of (U0.5Am0.5)O2-x that are responsible for the reaction could be explained by a binding model which is based on an americium valency state of + 3 and U5+. The existence of both valency states could be proved by XPS measurements. Due to the similar behaviour of neptunium and uranium the problems that are expected for the recycling of Np are much smaller than for americium

  7. Chemical aspects of the precise and accurate determination of uranium and plutonium from nuclear fuel solutions

    International Nuclear Information System (INIS)

    A method for the simultaneous or separate determination of uranium and plutonium has been developed. The method is based on the sorption of uranium and plutonium as their chloro complexes on Dowex 1x10 column. When separate uranium and plutonium fractions are desired, plutonium ions are reduced to Pu (III) and eluted, after which the uranium ions are eluted with dilute HCl. Simultaneous stripping of a mass ratio U/Pu approximately 1 fraction for mass spectrometric measurements is achieved by proper choice of eluant HC1 concentration. Special attention was paid to the obtaining of americium free plutonium fractions. The distribution coefficient measurements showed that at 12.5-M HCl at least 30 % of americium ions formed anionic chloro complexes. The chemical aspects of isotopic fractionation in a multiple filament thermal ionization source were also investigated. Samples of uranium were loaded as nitrates, chlorides, and sulphates and the dependence of the measured uranium isotopic ratios on the chemical form of the loading solution as well as on the filament material was studied. Likewise the dependence of the formation of uranium and its oxide ions on various chemical and instrumental conditions was investigated using tungsten and rhenium filaments. Systematic errors arising from the chemical conditions are compared with errors arising from the automatic evaluation of of spectra. (author)

  8. Influence of environmental factors on the gastrointestinal absorption of plutonium and americium

    International Nuclear Information System (INIS)

    The absorption of plutonium and americium from the gastrointestinal tract was studied, using adult hamsters and rabbits. Both actinides were administered as inorganic compounds, as organic complexes with naturally occurring chelating agents, and in a biologically incorporated form in liver tissues. The absorption of the tetravalent and hexavalent forms of plutonium were compared and the effect of protracted administration at very low concentrations was investigated. In addition, plutonium uptake from contaminated sediments and grass, collected near a nuclear-fuel reprocessing plant, was measured. The results of these studies suggest that chronic exposure of man to plutonium and americium in food and water will not lead to any substantial increase in their gastrointestinal absorption above the values currently recommended by the International Commission on Radiological Protection to define the occupational exposure of workers

  9. Speciation and bioavailability of plutonium and americium in the Irish Sea and other marine ecosystems

    International Nuclear Information System (INIS)

    Since the late 1960s, the Irish Sea has become a repository for a variety of radio-elements originating mainly in discharges from the British Nuclear Fuels (BNF) plc. Sellafield reprocessing complex located on the Cumbrian coast. In particular, transuranium nuclides such as plutonium, americium and curium (the main constituents of the α-emitting discharges) have become incorporated into every marine compartment by a variety of mechanisms, many of which are not well understood. Although extensive studies have been carried out in the near-field (eastern Irish Sea, especially in the vicinity of the discharge point and collateral muddy sediments), comparatively little had been done to assess the long-term behaviour and bioavailability of plutonium and americium in the far-field, e.g., the western Irish Sea, prior to the present study. In this dissertation, the results of an extensive research programme, undertaken in order to improve and refine our understanding of the behaviour of plutonium and americium in the marine environment, are presented. Specifically, the thesis details the results of (and conclusions deduced from) a series of experiments in which the physical and chemical speciation, colloidal association, mobility and bioavailability of plutonium and americium were examined in diverse environments including the Irish Sea and the Mediterranean. (author)

  10. Evaluation of the readsorption of plutonium and americium in dynamic fractionations of environmental solid samples

    DEFF Research Database (Denmark)

    Petersen, Roongrat; Hou, Xiaolin; Hansen, Elo Harald

    2008-01-01

    extractions. The degree of readsorption in dynamic and conventional batch extraction systems are compared and evaluated by using a double-spiking technique. A high degree of readsorption of plutonium and americium (>75%) was observed in both systems, and they also exhibited similar distribution patterns...

  11. Determination of specific activity of americium and plutonium in selected environmental samples

    International Nuclear Information System (INIS)

    The aim of this work was development of method for determination of americium and plutonium in environmental samples. Developed method was evaluated on soil samples and after they was applied on selected samples of fishes (smoked mackerel, herring and fillet from Alaska hake). The method for separation of americium is based on liquid separation with Aliquate-336, precipitation with oxalic acid and using of chromatographic material TRU-SpecTM.The intervals of radiochemical yields were from 13.0% to 80.9% for plutonium-236 and from 10.5% to 100% for americium-241. Determined specific activities of plutonium-239,240 were from (2.3 ± 1.4) mBq/kg to (82 ± 29) mBq/kg, the specific activities of plutonium-238 were from (14.2 ± 3.7) mBq/kg to (708 ± 86) mBq/kg. The specific activities of americium-241 were from (1.4 ± 0.9) mBq/kg to (3360 ± 210) mBq/kg. The fishes from Baltic Sea as well as from North Sea show highest specific activities then fresh-water fishes from Slovakia. Therefore the monitoring of alpha radionuclides in foods imported from territories with nuclear testing is recommended

  12. Determination of α-emitters (plutonium, americium, curium ...) in feces and urine ashes

    International Nuclear Information System (INIS)

    A description is given of the methods used to determine a number of radionuclides to be found in feces and urine, and obtain samples thin enough for counting and α-spectrometry. These methods can be applied to plutonium, americium and curium especially

  13. Comparison of radiotoxicity of uranium, plutonium, and thorium spent nuclear fuel at long-term storage

    International Nuclear Information System (INIS)

    Time dependence of radio-toxicity of actinides from spent uranium, MOX-plutonium, and thorium fuel calculated for storage during 1000 years is discussed in the paper. Calculations are based on the nuclear fuel of the VVER-1000 type reactor. Recommendations for uranium and plutonium spent fuel could be done to perform chemical separation of plutonium, americium, curium before long-term controllable storage. Americium should be separated after 50-70 years of storage for sufficient conversion of Pu-241 in Am-241. Cm-244 decays almost completely after 100 years. Extracted americium (possibly, with long-lived curium isotopes) should be directed to transmutation and plutonium should be reused. The separation of actinides is also effective to reduce decay heat power. In thorium spent fuel, the overwhelming share of radio-toxicity is determined by U-232. It is obvious that the repeated use of thorium fuel will be accompanied by accumulation of radio-toxicity. For a one-fold use of thorium fuel with deep U-233 burnup, it is necessary to perform additional deep burn-out (transmutation) of uranium fraction containing both U-233 and U-232. The further reduction of radio-toxicity by several orders can be obtained by extraction and transmutation of plutonium fraction (Pu-238). The transmutation of Th-228 - daughter nuclide of U-232 - is not necessary because Th-228 decays practically completely in 10 years together with its short-lived daughter nuclides

  14. An economic analysis of a light and heavy water moderated reactor synergy: burning americium using recycled uranium

    International Nuclear Information System (INIS)

    An economic analysis is presented for a proposed synergistic system between 2 nuclear utilities, one operating light water reactors (LWR) and another running a fleet of heavy water moderated reactors (HWR). Americium is partitioned from LWR spent nuclear fuel (SNF) to be transmuted in HWRs, with a consequent averted disposal cost to the LWR operator. In return, reprocessed uranium (RU) is supplied to the HWRs in sufficient quantities to support their operation both as power generators and americium burners. Two simplifying assumptions have been made. First, the economic value of RU is a linear function of the cost of fresh natural uranium (NU), and secondly, plutonium recycling for a third utility running a mixed oxide (MOX) fuelled reactor fleet has been already taking place, so that the extra cost of americium recycling is manageable. We conclude that, in order for this scenario to be economically attractive to the LWR operator, the averted disposal cost due to partitioning americium from LWR spent fuel must exceed 214 dollars per kg, comparable to estimates of the permanent disposal cost of the high level waste (HLW) from reprocessing spent LWR fuel. (authors)

  15. Plutonium and americium in arctic waters, the North Sea and Scottish and Irish coastal zones

    DEFF Research Database (Denmark)

    Hallstadius, L.; Aarkrog, Asker; Dahlgaard, Henning;

    1986-01-01

    Plutonium and americium have been measured in surface waters of the Greenland and Barents Seas and in the northern North Sea from 1980 through 1984. Measurements in water and biota, Fucus, Mytilus and Patella, were carried out in North-English and Scottish waters in 1982 and Fucus samples were...... plutonium from the latter to Spitsbergen waters. Fallout plutonium in Arctic waters has a residence time of the order of several years, while for Pu from Sellafield we estimate mean residence times of 11–15 months in Scottish waters and, tentatively, 1·5-3 y during transport from the North Channel (north...

  16. Plutonium and americium in arctic waters, the North Sea and Scottish and Irish coastal zones

    DEFF Research Database (Denmark)

    Hallstadius, L.; Aarkrog, Asker; Dahlgaard, Henning; Holm, E.; Boelskifte, S.; Duniec, S.; Persson, B.

    1986-01-01

    Plutonium and americium have been measured in surface waters of the Greenland and Barents Seas and in the northern North Sea from 1980 through 1984. Measurements in water and biota, Fucus, Mytilus and Patella, were carried out in North-English and Scottish waters in 1982 and Fucus samples were...... plutonium from the latter to Spitsbergen waters. Fallout plutonium in Arctic waters has a residence time of the order of several years, while for Pu from Sellafield we estimate mean residence times of 11–15 months in Scottish waters and, tentatively, 1·5-3 y during transport from the North Channel (north of...

  17. Effect of radiolysis on leachability of plutonium and americium from 76-101 glass. [Glass containing 2 mole % plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Nash, K.L.; Fried, S.; Friedman, A.M.; Susak, N.; Rickert, P.; Sullivan, J.C.; Karim, D.P.; Lam, D.J.

    1982-01-01

    One aspect of the leachability of actinide-bearing glass which has not been adequately addressed is the effect of radiolysis of the system (glass-water) on the amount of actinides liberated from the glass. In the present study, we have investigated the leaching of plutonium and americium from 76-101 glass samples (containing 2 mole % plutonium) in the presence of a one megaRad/hour gamma-radiation field. The presence of the radiation field was found to increase the leaching rate of both plutonium and americium by a factor of five. Speciation studies of the plutonium in the leachate indicate that the plutonium is present predominantly in the higher oxidation states, Pu(V) and Pu(VI) and that it is significantly associated with colloidal particles. Examination of the glass surfaces with x-ray photoemission spectroscopy, XPS, both before and after leaching was carried out; these studies showed lower surface concentrations of plutonium in the samples of glass leached in the radiation field. 1 figure, 3 tables.

  18. SKIN DOSIMETRY IN CONDITIONS OF ITS CONSTANT SURFACE CONTAMINATION WITH SOLUTIONS OF PLUTONIUM-239 AND AMERICIUM-241

    Directory of Open Access Journals (Sweden)

    E. B. Ershov

    2012-01-01

    Full Text Available The article considers, on the basis of experimental data, the issue of assessing dose burdens to the skin basal layer in conditions of its permanent contamination with solutions of plutonium-239 and americium-241 and subsequent decontamination.

  19. Isotopic and elemental composition of plutonium/americium oxides influence pulmonary and extra-pulmonary distribution after inhalation in rats

    International Nuclear Information System (INIS)

    The biodistribution of plutonium and americium has been studied in a rat model after inhalation of two PuO2 powders in lungs and extra-pulmonary organs from 3 d to 3 mo. The main difference between the two powders was the content of americium (approximately 46% and 4.5% of total alpha activity). The PuO2 with a higher proportion of americium shows an accelerated transfer of activity from lungs to blood as compared to PuO2 with the lower americium content, illustrated by increased urinary excretion and higher bone and liver actinide retention. The total alpha activity measured reflects mostly the americium biological behavior. The activity contained in epithelial lining fluid, recovered in the acellular phase of broncho-alveolar lavages, mainly contains americium, whereas plutonium remains trapped in macrophages. Epithelial lining fluid could represent a transitional pulmonary compartment prior to translocation of actinides to the blood and subsequent deposition in extra-pulmonary retention organs. In addition, differential behaviors of plutonium and americium are also observed between the PuO2 powders with a higher dissolution rate for both plutonium and americium being obtained for the PuO2 with the highest americium content. Our results indicate that the biological behavior of plutonium and americium after translocation into blood differ two-fold: (1) for the two actinides for the same PuO2 aerosol, and (2) for the same actinide from the two different aerosols. These results highlight the importance of considering the specific behavior of each contaminant after accidental pulmonary intake when assessing extra-pulmonary deposits from the level of activity excreted in urine or for therapeutic strategy decisions. (authors)

  20. Monte Carlo modeling of spallation targets containing uranium and americium

    International Nuclear Information System (INIS)

    Neutron production and transport in spallation targets made of uranium and americium are studied with a Geant4-based code MCADS (Monte Carlo model for Accelerator Driven Systems). A good agreement of MCADS results with experimental data on neutron- and proton-induced reactions on 241Am and 243Am nuclei allows to use this model for simulations with extended Am targets. It was demonstrated that MCADS model can be used for calculating the values of critical mass for 233,235U, 237Np, 239Pu and 241Am. Several geometry options and material compositions (U, U + Am, Am, Am2O3) are considered for spallation targets to be used in Accelerator Driven Systems. All considered options operate as deep subcritical targets having neutron multiplication factor of k∼0.5. It is found that more than 4 kg of Am can be burned in one spallation target during the first year of operation

  1. Contemporary state of plutonium and americium in the soils of Palesse state radiation-ecological reserve

    International Nuclear Information System (INIS)

    Full text: At present, the most important alpha-emitting radionuclides of Chernobyl origin are Pu 238, Pu 239, Pu 240 and Am 241. They are classified as the most dangerous group of radionuclides in view of the long half-lives and high radiotoxicity. The main part of alpha-emitted radionuclides is located within the Palesse State Radiation-Ecological Reserve. One of the most important factors determining the radioecological situation in the contaminated ecosystems is the physicochemical forms of radionuclides in a soil medium. Radionuclide species determine the radionuclide entrance into the soil solutions, their redistribution in soil profiles and the 'soil - plant' and the 'soil - surface, ground or underground water' systems as well as spreading beyond the contaminated area. The present work is devoted to investigation of state and migration ability of plutonium and americium in soils of the Palesse state radiation-ecological reserve after more than 20 years from the Chernobyl accident. The objects of investigation were mineral and organic soils sampled in 2008 with the step of 5 cm to the depth of 25-30 cm. The forms of plutonium and americium distinguishing by association with the different components of soil and by potential for migration in the soil medium were studied using the method of sequential selective extraction according to the modified Tessier scheme. Activities of Pu 238, Pu 239, Pu 240 and Am 241 in the samples were determined by the method of radiochemical analysis with alpha-spectrometer radionuclide identification. The dominant part of plutonium and americium in the soils is in immobile forms. Nowadays, radionuclide portions in water soluble and reversibly bound forms do not exceed 9.4 % of radionuclide content in the soil. In mineral soil samples, the radionuclide portions in these fractions exceed the corresponding portions in organic ones. In both mineral and organic soils, the portions of mobile americium are higher than plutonium. The

  2. Standard test method for quantitative determination of americium 241 in plutonium by Gamma-Ray spectrometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1994-01-01

    1.1 This test method covers the quantitative determination of americium 241 by gamma-ray spectrometry in plutonium nitrate solution samples that do not contain significant amounts of radioactive fission products or other high specific activity gamma-ray emitters. 1.2 This test method can be used to determine the americium 241 in samples of plutonium metal, oxide and other solid forms, when the solid is appropriately sampled and dissolved. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  3. Recovery of americium-241 from aged plutonium metal

    International Nuclear Information System (INIS)

    After separation and purification, both actinides were precipitated as oxalates and calcined. A large-scale process was developed using dissolution, separation, purification, precipitation, and calcination. Efforts were made to control corrosion, to avoid product contamination, to keep the volume of process and waste solutions manageable, and to denitrate solutions with formic acid. The Multipurpose Processing Facility (MPPF), designed for recovery of transplutonium isotopes, was used for the first time for the precipitation and calcination of americium. Also, for the first time,, large-scale formic acid denitration was performed in a canyon vessel at SRP

  4. Fabrication of neptunium, plutonium, americium and curium metals for fuel research

    International Nuclear Information System (INIS)

    The techniques for the fabrication of actinide metals; neptunium, americium and curium called as minor actinides, and plutonium, are surveied in a viewpoint of the preparation of starting materials for fuel property measurements. In this report, the processes of the conversion to metals, purification et al. are reviewed. The concept related to the apparatus design is also proposed and the considerable subjects are discussed. (author)

  5. Surface electronic structure and chemisorption of plutonium and uranium

    International Nuclear Information System (INIS)

    This paper reports that for the early actinides, uranium through plutonium, the direct overlap of the 5f orbitals and their hybridization with the s and d orbital give rise to narrow banding behavior for the 5f electrons. For the heavier actinides (from americium onward) the 5f electrons of the elemental solids remain localized and do not form a band. As the proportion of 5f to 6d and 7s electrons increases, the itinerant behavior of the 5f electrons gradually disappears. The change to localization is more dramatic between plutonium and americium. Surface f electrons tend to be somewhat more localized than those in the interior because of decreased opportunity for overlap and hybridization. It follows that the degree of localized f-electron behavior will be more conspicuous at the surface for any of the light actinides. Several experimentally accessible surface properties such as the work function are quite sensitive monitors of the degree of f-electron localization

  6. Comparison of acid leachate and fusion methods to determine plutonium and americium in environmental samples

    International Nuclear Information System (INIS)

    The Analytical Chemistry Laboratory at Argonne National Laboratory performs radiochemical analyses for a wide variety of sites within the Department of Energy complex. Since the chemical history of the samples may vary drastically from site to site, the effectiveness of any analytical technique may also vary. This study compares a potassium fluoride-pyrosulfate fusion technique with an acid leachate method. Both normal and high-fired soils and vegetation samples were analyzed for both americium and plutonium. Results show both methods work well, except for plutonium in high-fired soils. Here the fusion method provides higher accuracy

  7. Further Studies of Plutonium and Americium at Thule, Greenland

    DEFF Research Database (Denmark)

    Aarkrog, Asker; Dahlgaard, Henning; Nilsson, Karen Kristina;

    1984-01-01

    Eleven years after the accidental loss of nuclear weapons in 1968, the fourth scientific expedition to Thule occurred. The estimated inventory of 1 TBq 239,240Pu in the marine sediments was unchanged when compared with the estimate based on the 1974 data. Plutonium from the accident had moved...

  8. Liquid-liquid extraction separation and sequential determination of plutonium and americium in environmental samples by alpha-spectrometry

    International Nuclear Information System (INIS)

    A procedure is described by which plutonium and americium can be determined in environmental samples. The sample is leached with nitric acid and hydrogen peroxide, and the two elements are co-precipitated with ferric hydroxide and calcium oxalate. The calcium oxalate is incinerated at 4500 and the ash is dissolved in nitric acid. Plutonium is extracted with tri-n-octylamine solution in xylene from 4M nitric acid and stripped with ammonium iodide/hydrochloric acid. Americium is extracted with thenoyltrifluoroacetone solution in xylene at pH 4 together with rare-earth elements and stripped with 1M nitric acid. Americium and the rare-earth elements thus separated are sorbed on Dowex 1 x 4 resin from 1M nitric acid in 93% methanol, the rare-earth elements are eluted with 0.1M hydrochloric acid/0.5M ammonium thiocyanate/80% methanol and the americium is finally eluted with 1.5M hydrochloric acid in 86% methanol. Plutonium and americium in each fraction are electro-deposited and determined by alpha-spectrometry. Overall average recoveries are 81% for plutonium and 59% for americium. (author)

  9. Comparative study of plutonium and americium bioaccumulation from two marine sediments contaminated in the natural environment

    International Nuclear Information System (INIS)

    Plutonium and americium sediment-animal transfer was studied under controlled laboratory conditions by exposure of the benthic polychaete Nereis diversicolor (O. F. Mueller) to marine sediments contaminated by a nuclear bomb accident (near Thule, Greenland) and nuclear weapons testing (Enewetak Atoll). In both sediment regimes, the bioavailability of plutonium and 241Am was low, with specific activity in the tissues 241Am occurred and 241Am uptake from the Thule sediment was enhanced compared to that from lagoon sediments of Enewetak Atoll. Autoradiography studies indicated the presence of hot particles of plutonium in the sediments. The results highlight the importance of purging animals of their gut contents in order to obtain accurate estimates of transuranic transfer from ingested sediments into tissue. It is further suggested that enhanced transuranic uptake by some benthic species could arise from ingestion of highly activity particles and organic-rich detritus present in the sediments. (author)

  10. Standard practice for The separation of americium from plutonium by ion exchange

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This practice describes the use of an ion exchange technique to separate plutonium from solutions containing low concentrations of americium prior to measurement of the 241Am by gamma counting. 1.2 This practice covers the removal of plutonium, but not all the other radioactive isotopes that may interfere in the determination of 241Am. 1.3 This practice can be used when 241Am is to be determined in samples in which the plutonium is in the form of metal, oxide, or other solid provided that the solid is appropriately sampled and dissolved (See Test Methods C758, C759, and C1168). 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  11. Plutonium and americium behavior in coral atoll environments

    International Nuclear Information System (INIS)

    Inventories of 239+240Pu and 241Am greatly in excess of global fallout levels persist in the benthic environments of Bikini and Enewetak Atolls. Quantities of 239+240Pu and lesser amounts of 241Am are continuously mobilizing from these sedimentary reservoirs. The amount of 239+240Pu mobilized to solution at any time represents 0.08 to 0.09% of the sediment inventories to a depth of 16 cm. The mobilized 239+240Pu has solute-like characteristics and different valence states coexist in solution - the largest fraction of the soluble plutonium is in an oxidized form (+V,VI). The adsorption of plutonium to sediments is not completely reversible because of changes that occur in the relative amounts of the mixed oxidation states in solution with time. Further, any characteristics of 239+240Pu described at one location may not necessarily be relevant in describing its behavior elsewhere following mobilization and migration. The relative amounts of 241Am to 239+240Pu in the sedimentary deposits at Enewetak and Bikini may be altered in future years because of mobilization and radiological decay. Mobilization of 239+240Pu is not a process unique to these atolls, and quantities in solution derived from sedimentary deposits can be found at other global sites. These studies in the equatorial Pacific have significance in assessing the long-term behavior of the transuranics in any marine environment. 22 references, 1 figure, 13 tables

  12. Determination of trace concentration of uranium in americium oxide samples by ICP-AES

    International Nuclear Information System (INIS)

    A solvent extraction method has been developed for the determination of uranium (200-2000 ppm) in americium oxide samples. The method involves the quantitative separation of uranium from americium matrix using mixed solvent comprising 1.1M tri-n-butyl phosphate (TBP) +1% trialkyl phosphine oxide (TRPO) + 0.3 M tertiary butyl hydroquinone (TBHQ) in n-dodecane. Uranium from the organic is stripped into the aqueous phase with 0.8 M oxalic acid and determined by ICP-AES. The reliability of the method was ascertained by analytical recovery, which is found to be nearly 100%. (author)

  13. Transfer of radiocaesium, plutonium and americium to sheep after ingestion of contaminated soil

    International Nuclear Information System (INIS)

    A dual isotope method has been used to study the transfer of 137Cs, 239/240Pu and 241Am to sheep following ingestion of contaminated soil. Two soils were used; an alluvial gley contaminated by Sellafield discharges, and an organic soil, artificially contaminated in a lysimeter. Values of the true absorption coefficient of radiocaesium of 0.19 +/- 0.03 and 0.03 +/- 0.01 respectively were obtained for these soils. This implies an availability factor for soil-associated radiocaesium of up to about 20 pc compared to radiocaesium ingested in soluble form. The absorption of plutonium and americium was not significantly different for the two soils tested. Absorption of both plutonium and americium was in the range 10-5 - 10-4, with mean values of 7 x 10-5 and 4 x 10-5 obtained respectively. These values imply availability factors of around 10 pc, compared to the value of 5 x 10-4 recommended by ICRP for plutonium ingested in a comparatively available form. These results are compared with estimates of availability made using an in-vitro approach

  14. Transfer of radiocaesium, plutonium and americium to sheep after ingestion of contaminated soil

    Energy Technology Data Exchange (ETDEWEB)

    Cooke, A.I.; Weekes, T.E.C. [Newcastle upon Tyne Univ. (United Kingdom). Dept. of Biological and Nutritional Sciences; Rimmer, D.L. [Newcastle upon Tyne Univ. (United Kingdom). Dept. of Agricultural and Environmental Science; Green, N.; Wilkins, B.T. [National Radiological Protection Board, Chilton (United Kingdom)

    1997-12-31

    A dual isotope method has been used to study the transfer of {sup 137}Cs, {sup 239/240}Pu and {sup 241}Am to sheep following ingestion of contaminated soil. Two soils were used; an alluvial gley contaminated by Sellafield discharges, and an organic soil, artificially contaminated in a lysimeter. Values of the true absorption coefficient of radiocaesium of 0.19 +/- 0.03 and 0.03 +/- 0.01 respectively were obtained for these soils. This implies an availability factor for soil-associated radiocaesium of up to about 20 pc compared to radiocaesium ingested in soluble form. The absorption of plutonium and americium was not significantly different for the two soils tested. Absorption of both plutonium and americium was in the range 10{sup -5} - 10{sup -4}, with mean values of 7 x 10{sup -5} and 4 x 10{sup -5} obtained respectively. These values imply availability factors of around 10 pc, compared to the value of 5 x 10{sup -4} recommended by ICRP for plutonium ingested in a comparatively available form. These results are compared with estimates of availability made using an in-vitro approach

  15. Report of scouting study on precipitation of strontium, plutonium, and americium from Hanford complexant concentrate waste

    International Nuclear Information System (INIS)

    A laboratory scouting test was conducted of precipitation methods for reducing the solubility of radionuclides in complexant concentrate (CC) waste solution. The results show that addition of strontium nitrate solution is effective in reducing the liquid phase activity of 90Sr (Strontium) in CC waste from tank 107-AN by 94% when the total strontium concentration is adjusted to 0.1 M. Addition of ferric nitrate solution effective in reducing the 241Am (Americium) activity in CC waste by 96% under the conditions described in the report. Ferric nitrate was also marginally effective in reducing the solubility of 239/240Pu (Plutonium) in CC waste

  16. Purification of used scintillation liquids containing the alpha emitters americium and plutonium

    International Nuclear Information System (INIS)

    In Sweden, alpha radioactive waste liquids with an activity over some kBq per waste container cannot be sent for final storage. Therefore, in this work, a method for a purification of alpha active scintillation cocktails was developed. Until today (March, 2013) more than 20 L of scintillation liquids have successfully been purified from americium and plutonium. The products of the process are a solid fraction that can be sent to final storage and a practically non-radioactive liquid fraction that can be sent to municipal incineration. (author)

  17. Experimental and in situ investigations on americium, curium and plutonium behaviour in marine benthic species: transfer from water or sediments

    International Nuclear Information System (INIS)

    The tranfer of transuranic elements -americium, curium and plutonium- from the sediments containing them to some marine benthic species (endofauna and epifauna) was studied with a twofold approach - laboratory and in-situ investigation. The experimental investigations, divided into three parts, made it possible to specify concentration factors (F.C.), transfer factors (F.T.) and to understand the process involved for 5 benthic species. The result were refined by an in-situ study that brought new data on the marine distribution of the transuranic elements released by the La Hague plant. Finally, the localization of americium and plutonium in the tissues and cells of these species was determined by autoradiography

  18. Direct isotope ratio analysis of individual uranium-plutonium mixed particles with various U/Pu ratios by thermal ionization mass spectrometry.

    Science.gov (United States)

    Suzuki, Daisuke; Esaka, Fumitaka; Miyamoto, Yutaka; Magara, Masaaki

    2015-02-01

    Uranium and plutonium isotope ratios in individual uranium-plutonium (U-Pu) mixed particles with various U/Pu atomic ratios were analyzed without prior chemical separation by thermal ionization mass spectrometry (TIMS). Prior to measurement, micron-sized particles with U/Pu ratios of 1, 5, 10, 18, and 70 were produced from uranium and plutonium certified reference materials. In the TIMS analysis, the peaks of americium, plutonium, and uranium ion signals were successfully separated by continuously increasing the evaporation filament current. Consequently, the uranium and plutonium isotope ratios, except the (238)Pu/(239)Pu ratio, were successfully determined for the particles at all U/Pu ratios. This indicates that TIMS direct analysis allows for the measurement of individual U-Pu mixed particles without prior chemical separation. PMID:25479434

  19. Dissolution of uranium and plutonium oxide using TBP-HNO3 complex. Research document

    International Nuclear Information System (INIS)

    The current technology for the selective separation of plutonium and uranium from spent nuclear fuel (MOX) using TBP-HNO3 complex is being developed (Powdered fuel extraction process). It is promising to simplify the reprocessing process for the selective separation because of its potential to unite the chemical processes, dissolution process using nitric acid and co-extraction process using TBP solvent, and to operate under the ambient pressure and at relatively 'mild' temperature. Plutonium oxide has reported to provide slower dissolution than uranium oxide in nitric acid. In this work dissolution behaviors of plutonium into TBP-HNO3 complex from powdered plutonium and uranium mixed oxide were examined. The powdered MOX fuel (average particles size 10 μm) was prepared from PuO2-UO2 pellets by heating for 4 hours at 400degC. The prepared powder was dissolved into TBP-4.74mol/L HNO3 complex and was stirred for 300 minutes. In the test with 6 grams of powdered MOX fuel and 20 mL of the TBP-HNO3 complex, the concentration of plutonium reached 0.17 mol/L and about 90 percent of plutonium was dissolved. It is experimentally confirmed plutonium was dissolved into the TBP-HNO3 complex from plutonium and uranium mixed oxide. The early dissolution rate was almost the same as that obtained with nitric acid solution. It is likely to predict the dissolution rate from the rate for nitric acid solution. Americium that was contained in the MOX fuel was also dissolved into the TBP-HNO3 complex, but was slower than plutonium. (author)

  20. Comparative study of plutonium and americium bioaccumulation from two marine sediments contaminated in the natural environment

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, T.F.; Smith, J.D. (Melbourne Univ., Parkville (Australia). Dept. of Inorganic Chemistry); Fowler, S.W.; LaRosa, J.; Holm, E. (International Atomic Energy Agency, Monaco-Ville (Monaco). Lab. of Marine Radioactivity); Aarkrog, A.; Dahlgaard, H. (Risoe National Lab., Roskilde (Denmark))

    1991-01-01

    Plutonium and americium sediment-animal transfer was studied under controlled laboratory conditions by exposure of the benthic polychaete Nereis diversicolor (O. F. Mueller) to marine sediments contaminated by a nuclear bomb accident (near Thule, Greenland) and nuclear weapons testing (Enewetak Atoll). In both sediment regimes, the bioavailability of plutonium and {sup 241}Am was low, with specific activity in the tissues <1% (dry wt) than in the sediments. Over the first three months, a slight preference in transfer of plutonium over {sup 241}Am occurred and {sup 241}Am uptake from the Thule sediment was enhanced compared to that from lagoon sediments of Enewetak Atoll. Autoradiography studies indicated the presence of hot particles of plutonium in the sediments. The results highlight the importance of purging animals of their gut contents in order to obtain accurate estimates of transuranic transfer from ingested sediments into tissue. It is further suggested that enhanced transuranic uptake by some benthic species could arise from ingestion of highly activity particles and organic-rich detritus present in the sediments. (author).

  1. Purification of scintillation cocktails containing the alpha emitters americium and plutonium

    International Nuclear Information System (INIS)

    One efficient way of measuring alpha emitters is by the usage of liquid scintillation counting (LSC). A liquid sample is placed in a vial containing a scintillation cocktail. The alpha particles excite electrons in the surrounding liquid, and when they are de-excited photons are emitted. The photons are detected and the activity can be quantified. LSC has a high efficiency for alpha radiation and is therefore a fast and easy way for measuring alpha emitting samples. One drawback is that it does not differentiate very well between alpha energies; measurements of for example curium and plutonium simultaneously are impossible and demand other techniques. Another drawback is the production of a liquid alpha active waste. In Sweden alpha radioactive waste liquids with an activity over some kBq per waste container cannot be sent for final storage. If, however, the activity of the liquids could be reduced by precipitation of the actinides, it would be possible to send away the liquid samples to municipal incineration. In this work a method for a purification of alpha active scintillation cocktails was developed. The method was first tried on a lab scale, and then scaled up. Until today (March, 2013) more than 20 liters of scintillation liquids have successfully been purified from americium and plutonium at Chalmers University of Technology in Sweden. The four scintillation cocktails used were Emulsifier Safe®, Hionic-Fluor®, Ultima Gold AB® and Ultima Gold XR®. The scintillation cocktails could all be purified from americium with higher yield than 95%. The yield was kept when the liquids were mixed. Also plutonium could be precipitated with a yield over 95% in all cocktails except in Hionic-Fluor® (>55%). However, that liquid in particular could be purified (>95%) by mixing it with the three other cocktails. Up-scaling was performed to a batch size of 6-8 L of scintillation cocktail. In neither the americium nor the plutonium system, adverse effects of increasing the

  2. Effects of Hanford high-level waste components on sorption of cobalt, strontium, neptunium, plutonium, and americium on Hanford sediments

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, C H; Barney, G S

    1983-03-01

    To judge the feasibility of continued storage of high-level waste solutions in existing tanks, effects of chemical waste components on the sorption of hazardous radioelements were determined. Experiments identified the effects of 12 Hanford high-level waste-solution components on the sorption of cobalt, strontium, neptunium, plutonium, and americium on 3 Hanford 200 Area sediments. The degree of sorption of strontium, neptunium, plutonium, and americium on two Hanford sediments was then quantified in terms of the concentrations of the influential waste components. Preliminary information on the influence of the waste components on radioelement solubility was gathered. Of the 12 Hanford waste-solution components studied, the most influential on radioelement sorption were NaOH, NaAlO/sub 2/, HEDTA, and EDTA. The chelating complexants, HEDTA and EDTA, generally decreased sorption by complexation of the radioelement metal ions. The components NaOH and NaAlO/sub 2/ decreased neptunium and plutonium sorption and increased cobalt sorption. Americium sorption was increased by NaOH. The three Hanford sediments' radioelement sorption behaviors were similar, implying that their sorption reactions were also similar. Sorption prediction equations were generated for strontium, neptunium, plutonium, and americium sorption reactions on two Hanford sediments. The equations yielded values of the distribution coefficient, K/sub d/, as quadratic functions of waste-component concentrations and showed that postulated radioelement migration rates through Hanford sediment could change by factors of 13 to 40 by changes in Hanford waste composition.

  3. Effects of Hanford high-level waste components on sorption of cobalt, strontium, neptunium, plutonium, and americium on Hanford sediments

    International Nuclear Information System (INIS)

    To judge the feasibility of continued storage of high-level waste solutions in existing tanks, effects of chemical waste components on the sorption of hazardous radioelements were determined. Experiments identified the effects of 12 Hanford high-level waste-solution components on the sorption of cobalt, strontium, neptunium, plutonium, and americium on 3 Hanford 200 Area sediments. The degree of sorption of strontium, neptunium, plutonium, and americium on two Hanford sediments was then quantified in terms of the concentrations of the influential waste components. Preliminary information on the influence of the waste components on radioelement solubility was gathered. Of the 12 Hanford waste-solution components studied, the most influential on radioelement sorption were NaOH, NaAlO2, HEDTA, and EDTA. The chelating complexants, HEDTA and EDTA, generally decreased sorption by complexation of the radioelement metal ions. The components NaOH and NaAlO2 decreased neptunium and plutonium sorption and increased cobalt sorption. Americium sorption was increased by NaOH. The three Hanford sediments' radioelement sorption behaviors were similar, implying that their sorption reactions were also similar. Sorption prediction equations were generated for strontium, neptunium, plutonium, and americium sorption reactions on two Hanford sediments. The equations yielded values of the distribution coefficient, K/sub d/, as quadratic functions of waste-component concentrations and showed that postulated radioelement migration rates through Hanford sediment could change by factors of 13 to 40 by changes in Hanford waste composition

  4. The distribution of plutonium-239 and americium-241 in the Syrian hamster following its intravenous administration as citrate

    International Nuclear Information System (INIS)

    Actinide distribution in various tissues and the skeleton of hamsters by liquid scintillation counting or isotope dilution. For plutonium 57% of activity was concentrated in the skeleton and more than 90% in the liver and skeleton after seven days. For americium the liver retained more than 50% of total activity and 25% was excreted in urine within seven days. (U.K.)

  5. Americium and plutonium in water, biota, and sediment from the central Oregon coast

    International Nuclear Information System (INIS)

    Plutonium-239, 240 and americium-241 were measured in the mussel Mytilus californianus from the region of Coos Bay, OR. The flesh of this species has a plutonium concentration of about 90 fCi/kg, and an Am-241/Pu-239, 240 ratio that is high relative to mixed fallout, ranging between two and three. Transuranic concentrations in sediment, unfiltered water, and filterable particulates were also measured; none of these materials has an Am/Pu ratio as greatly elevated as the mussels, and there is no apparent difference in the Am/Pu ratio of terrestrial runoff and coastal water. Sediment core profiles do not allow accumulation rates or depositional histories to be identified, but it does not appear that material characterized by a high Am/Pu ratio has ever been introduced to this estuary. Other bivalves (Tresus capax and Macoma nasuta) and a polychaete (Abarenicola sp.) do not have an elevated Am/Pu ratio, although the absolute activity of plutonium in the infaunal bivalves is roughly four times that in the mussels

  6. Effect of a long-term release of plutonium and americium into an estuarine and coastal sea ecosystem

    International Nuclear Information System (INIS)

    This paper discusses the general problem of speciation of plutonium and americium in aquatic ecosystems and the implications relative to their fate in those systems. The following conclusions were reached: several oxidation states of plutonium coexist in the natural environment; the effect of environmental changes such as pH and Esub(h) values and complexes are probably the cause of these various oxidation states; a clearer definition of the 'concentration factor' should be given in view of the important role the sediments play in supplying plutonium for transfer through the food web. (author)

  7. Tags to Track Illicit Uranium and Plutonium

    International Nuclear Information System (INIS)

    With the expansion of nuclear power, it is essential to avoid nuclear materials from falling into the hands of rogue nations, terrorists, and other opportunists. This paper examines the idea of detection and attribution tags for nuclear materials. For a detection tag, it is proposed to add small amounts [about one part per billion (ppb)] of 232U to enriched uranium to brighten its radioactive signature. Enriched uranium would then be as detectable as plutonium and thus increase the likelihood of intercepting illicit enriched uranium. The use of rare earth oxide elements is proposed as a new type of 'attribution' tag for uranium and thorium from mills, uranium and plutonium fuels, and other nuclear materials. Rare earth oxides are chosen because they are chemically compatible with the fuel cycle, can survive high-temperature processing operations in fuel fabrication, and can be chosen to have minimal neutronic impact within the nuclear reactor core. The mixture of rare earths and/or rare earth isotopes provides a unique 'bar code' for each tag. If illicit nuclear materials are recovered, the attribution tag can identify the source and lot of nuclear material, and thus help police reduce the possible number of suspects in the diversion of nuclear materials based on who had access. (authors)

  8. Recovery of plutonium and americium from chloride salt wastes by solvent extraction

    International Nuclear Information System (INIS)

    Plutonium and americium can be recovered from aqueous waste solutions containing a mixture of HCl and chloride salt wastes by the coupling of two solvent extraction systems: tributyl phosphate (TBP) in tetrachloroethylene (TCE) and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) in TCE. In the flowsheet developed, the salt wastes are dissolved in HCl, the Pu(III) is oxidized to the IV state with NaClO2 and recovered in the TBP-TCE cycle, and the Am is then removed from the resultant raffinate by the CMPO-TCE cycle. The consequences of the feed solution composition and extraction behavior of these species on the process flowsheet design, the Pu-product purity, and the decontamination of the aqueous raffinate from transuranic elements are discussed. 16 refs., 6 figs

  9. Recovery of plutonium and americium from chloride salt wastes by solvent extraction

    International Nuclear Information System (INIS)

    Plutonium and americium can be recovered from aqueous waste solutions containing a mixture of HCl and chloride salt wastes by the coupling of two solvent extraction systems: tributyl phosphate (TBP) in tetrachloroethylene (TCE) and octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) in TCE. In the flowsheet developed, the salt wastes are dissolved in HCl, the Pu(III) is oxidized to the IV state with NaClO2 and recovered in the TBP-TCE cycle, and the Am is then removed from the resultant raffinate by the CMPO-TCE cycle. The consequences of the feed solution composition and extraction behavior of these species on the process flowsheet design, the Pu-product purity, and the decontamination of the aqueous raffinate from transuranic elements are discussed

  10. The uptake of plutonium-239, 240, americium-241, strontium-90 into plants

    International Nuclear Information System (INIS)

    This report describes the results of measurements on the uptake of plutonium, americium, strontium-90 and caesium-137 into peas, beet, oats, sweet corn, tomatoes and vegetable marrow grown in tubs containing radioactively-contaminated silts. The silts had been taken from an area of West Cumbria commonly referred to as the Ravenglass estuary. The experiments are categorised as being carried out under non-standard conditions because of the manner in which the radioactivity came to be incorporated into the growth medium. The growth medium was representative of conditions which could arise when the estuarine silt moves inland under the influence of wind and tide and mixes with the adjacent farm land. The silt had been contaminated by radioactive effluents from the nuclear fuels reprocessing plant at Sellafield and this contamination had been brought about by natural means. (Auth.)

  11. Ab initio modelling of the behaviour of helium in americium and plutonium oxides

    International Nuclear Information System (INIS)

    By means of an ab initio plane wave pseudo potential method, plutonium dioxide and americium dioxide are modelled, and the behaviour of helium in both these materials is studied. We first show that a pseudo potential approach in the Generalized Gradient Approximation (GGA) can satisfactorily describe the cohesive properties of PuO2 and AmO2. We then calculate the formation energies of point defects (vacancies and interstitials), as well as the incorporation and solution energies of helium in PuO2 and AmO2. The results are discussed according to the incorporation site of the gas atom in the lattice and to the stoichiometry of PuO2±x and AmO2±x. (authors)

  12. Removal of plutonium and americium from hydrochloric acid waste stream using extraction chromatography

    International Nuclear Information System (INIS)

    Extraction chromatography is under development as a method to lower actinide activity levels in hydrochloric acid (HCl) effluent streams. Successful application of this technique would allow recycle of the largest portion of HCl, while lowering the quantity and improving the form of solid waste generated. The extraction of plutonium and americium from HCl solutions was examined for several commercial and similar laboratory-produced resins coated with n-octyl(phenyl)-N,N-diisobutylcarbamoylmethyphosphine oxide (CMPO) and either tributyl phosphate (TBP), or diamyl amylphosphonate (DAAP). Distribution coefficients for Pu and Am were measured by contact studies in hydrochloric acid solutions over the range of 0.1 - 10.0 N HCl, whole varying REDOX conditions, actinide loading levels, and contact time intervals. Significant differences in the actinide distribution coefficients, and in the kinetics of actinide removal were observed as a function of resin formulation. The usefulness of these resins for actinide removal from HCl effluent streams is discussed

  13. Plutonium and americium concentrations and vertical profiles in some Italian mosses used as bioindicators

    International Nuclear Information System (INIS)

    We have examined the uptake of actinide elements Am and Pu by different species of lichen and moss collected in two locations (Urbino, Central Italy; Alps region, North-east Italy). Plutonium and americium were separated and determined by extraction chromatography, electrodeposition and alpha-spectrometry. This paper summarizes our results with a special emphasis on the vertical profiles of these actinides in two different species of mosses. Several 1-2 cm depth sections were obtained and dated by 210Pb method. A typical peak for 239,240Pu and 241Am was found in the very old moss species ('Sphagnum Compactum') at a depth corresponding to the period 1960-1970 which was the period characterized by the maximum nuclear weapon tests. In a younger moss species ('Neckeria Crispa') no peak was observed and the regression curves showed that Am is more mobile than 239,240Pu and 238Pu. (author)

  14. Determination of uranium and plutonium in high active solution by extractive spectrophotometry

    International Nuclear Information System (INIS)

    A method for determination of uranium and plutonium in high active solution by extractive spectrophotometry was developed. TOPO in xylene was used as extractant for uranium and plutonium from irradiated plutonium carbide and uranium carbide

  15. Consideration of the effect of lymph-node deposition upon the measurement of plutonium and americium in the lungs

    International Nuclear Information System (INIS)

    Measurement of an inhaled radionuclide by external photon counting includes quantities which may be contained in lymph nodes, as well as quantities in the lungs. An overestimate of the lung burden can result, if a portion of the radionuclide were present in the lymph nodes. This problem is analyzed with respect to the measurement of inhaled plutonium containing plutonium-241 and americium-241, when americium-241 has been used as a tracer for the plutonium. Equations are derived which yield the amounts of americium and of plutonium in the lungs and in the lymph nodes as a function of time after exposure and for various translocation and retention parameters. Count histories (count profiles) of actual exposure cases are compared with calculated count profiles in order to gain insight into possible values of the translocation and retention parameters. Comparison is also made with calculated count profiles using values of translocation and retention parameters recommended by the International Commission on Radiological Protection (ICRP) for use with the Task Group Lung Model. The magnitude of the possible overestimate (error factor) was calculated for combinations o

  16. Influence of biofilms on migration of uranium, americium and europium in the environment

    International Nuclear Information System (INIS)

    The report on the influence of biofilms on migration of uranium, americium and europium in the environment deals with the contamination problems of uranium mines such as SDAG WISMUT in Saxonia and Thuringia. In mine waters microorganisms form a complex microbiological biocoenosis in spite of low pH values and high heavy metal concentrations including high uranium concentrations. The analyses used microbiological methods like confocal laser scanning microscopy and molecular-biological techniques. The interactions of microorganism with fluorescent radioactive heavy metal ions were performed with TRLFS (time resolved laser-induced fluorescence spectroscopy).

  17. Combined radiochemical procedure for determination of plutonium, americium and strontium-90 in the soil samples from SNTS

    International Nuclear Information System (INIS)

    The results of combined radiochemical procedure for the determination of plutonium, americium and 90Sr (via measurement of 90Y) in the soil samples from SNTS (Semipalatinsk Nuclear Test Site) are presented. The processes of co-precipitation of these nuclides with calcium fluoride in the strong acid solutions have been investigated. The conditions for simultaneous separation of americium and yttrium using extraction chromatography have been studied. It follows from analyses of real soil samples that the procedure developed provides the chemical recovery of plutonium and yttrium in the range of 50-95 % and 60-95 %, respectively. The execution of the procedure requires 3.5 working days including a sample decomposition study. (author)

  18. Evaluation of uranium and plutonium internal standards

    International Nuclear Information System (INIS)

    Two internal standards have been synthesized and evaluated for application in mass spectrometric analysis of uranium. Each standard is a mixture of highly enriched 233U and 236U. One is used to refine 235U/238U ratio measurements and the other for isotope dilution applications. An internal standard consisting of a mixture of 242Pu and 244Pu has been characterized for use in plutonium analyses. Precisions obtainable on pulse-counting instruments have been improved to about 0.1%. 21 refs., 13 tabs

  19. A study of plutonium and americium concentrations in seaspray on the southern Scottish coast

    International Nuclear Information System (INIS)

    Seaspray and seawater have been collected from the southern Scottish coast and, for comparison, Cumbria in northwest England during 1989 and 1991. The occurrence of sea-to-land transfer of the actinides plutonium and americium in seaspray was observed on these coasts using muslin screens (a semi-quantitative technique most efficient for collecting large spray droplets) and high volume conventional air samplers. The actinides and fine particulate in the spray were present in relatively higher concentrations than measured in the adjacent seawater, i.e. the spray was enriched in particulate actinides. The net efficiency of the muslim screens in collecting airborne plutonium isotopes and 241Am generally appeared to be about 20%. A review of earlier published concentrations of 239+240Pu and 241Am measured in aerosol and deposition for over a year several tens of metres inland was carried out. This suggested that airborne activities are up to a factor of 5 times higher in Cumbria than southern Scotland. However, neither the new data collected in 1989 and 1991 nor this older data suggests any enhancement of seaspray actinide enrichment in southern Scotland compared to Cumbria. This finding contrasts with earlier, more limited, comparisons that have been carried out which suggested such a difference. There is clear evidence of considerable localised spatial and temporal variability in aerosol actinide enrichment over the beaches in both areas. Enrichments varies between 20 and 500 relative to the adjacent surf zone waters. However, the average enrichment in spray based on the continuous measurements made further inland is likely to be at the lower end of this range. (author)

  20. Sorption of plutonium and americium on repository, backfill and geological materials relevant to the JNFL low-level radioactive waste repository at Rokkasho-Mura

    International Nuclear Information System (INIS)

    An integrated program of batch sorption experiments and mathematical modeling has been carried out to study the sorption of plutonium and americium on a series of repository, backfill and geological materials relevant to the JNFL low-level radioactive waste repository at Rokkasho-Mura. The sorption of plutonium and americium on samples of concrete, mortar, sand/bentonite, tuff, sandstone and cover soil has been investigated. In addition, specimens of bitumen, cation and anion exchange resins, and polyester were chemically degraded. The resulting degradation product solutions, alongside solutions of humic and isosaccharinic acids were used to study the effects on plutonium sorption onto concrete, sand/bentonite and sandstone. The sorption behavior of plutonium and americium has been modeled using the geochemical speciation program HARPHRQ in conjunction with the HATCHES database

  1. Determination of plutonium americium and curium in soil samples by solvent extraction with trioctylphosphine oxide

    International Nuclear Information System (INIS)

    A method of Pu, Am and Cm determination in soil samples, which was developed for analyzing samples from territories subjected to radioactive contamination as a result of the Chernobyl accident is described. After preliminary treatment the samples were leached by solution of 7 mol/l HNO23+0.3 mol/l KBrO3 during heating. Pu was isolated by extraction with 0.05 mol TOPO from 7 mol/l HNO3. 144Ce and partially remaining in water phase isotopes of Zr, U and Th were isolated in an extraction-chromatographic column with TOPO and PbO2. Then Am and Cm were extracted by 0.2 mol/l TOPO from solution 1 mol/l HLact+0.07 mol/l DTPA+1 mol/l Al(NO3)3. Alpha-activity of both extracted products was determined in liquid scintillation counter. Chemical yield of plutonium counted to 85±10%, that of americium and curium -75±10%. 17 refs

  2. In Vitro Dissolution Tests of Plutonium and Americium Containing Contamination Originating From ZPPR Fuel Plates

    Energy Technology Data Exchange (ETDEWEB)

    William F. Bauer; Brian K. Schuetz; Gary M. Huestis; Thomas B. Lints; Brian K. Harris; R. Duane Ball; Gracy Elias

    2012-09-01

    Assessing the extent of internal dose is of concern whenever workers are exposed to airborne radionuclides or other contaminants. Internal dose determinations depend upon a reasonable estimate of the expected biological half-life of the contaminants in the respiratory tract. One issue with refractory elements is determining the dissolution rate of the element. Actinides such as plutonium (Pu) and Americium (Am) tend to be very refractory and can have biological half-lives of tens of years. In the event of an exposure, the dissolution rates of the radionuclides of interest needs to be assessed in order to assign the proper internal dose estimates. During the November 2011 incident at the Idaho National Laboratory (INL) involving a ZPPR fuel plate, air filters in a constant air monitor (CAM) and a giraffe filter apparatus captured airborne particulate matter. These filters were used in dissolution rate experiments to determine the apparent dissolution half-life of Pu and Am in simulated biological fluids. This report describes these experiments and the results. The dissolution rates were found to follow a three term exponential decay equation. Differences were noted depending upon the nature of the biological fluid simulant. Overall, greater than 95% of the Pu and 93% of the Am were in a very slow dissolving component with dissolution half-lives of over 10 years.

  3. Solution speciation of plutonium and Americium at an Australian legacy radioactive waste disposal site.

    Science.gov (United States)

    Ikeda-Ohno, Atsushi; Harrison, Jennifer J; Thiruvoth, Sangeeth; Wilsher, Kerry; Wong, Henri K Y; Johansen, Mathew P; Waite, T David; Payne, Timothy E

    2014-09-01

    During the 1960s, radioactive waste containing small amounts of plutonium (Pu) and americium (Am) was disposed in shallow trenches at the Little Forest Burial Ground (LFBG), located near the southern suburbs of Sydney, Australia. Because of periodic saturation and overflowing of the former disposal trenches, Pu and Am have been transferred from the buried wastes into the surrounding surface soils. The presence of readily detected amounts of Pu and Am in the trench waters provides a unique opportunity to study their aqueous speciation under environmentally relevant conditions. This study aims to comprehensively investigate the chemical speciation of Pu and Am in the trench water by combining fluoride coprecipitation, solvent extraction, particle size fractionation, and thermochemical modeling. The predominant oxidation states of dissolved Pu and Am species were found to be Pu(IV) and Am(III), and large proportions of both actinides (Pu, 97.7%; Am, 86.8%) were associated with mobile colloids in the submicron size range. On the basis of this information, possible management options are assessed. PMID:25126837

  4. Determination of Trace Plutonium in Uranium Product by ID-ICP-MS

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Plutonium is strictly limited in the uranium product of spent fuel reprocessing. The analysis of plutonium in uranium product is the key point of product quality control. Plutonium concentration is limited below

  5. Influence of biofilms on migration of uranium, americium and europium in the environment; Einfluss von Biofilmen auf das Migrationsverhalten von Uran, Americium und Europium in der Umwelt

    Energy Technology Data Exchange (ETDEWEB)

    Baumann, Nils; Zirnstein, Isabel; Arnold, Thuro

    2015-07-01

    The report on the influence of biofilms on migration of uranium, americium and europium in the environment deals with the contamination problems of uranium mines such as SDAG WISMUT in Saxonia and Thuringia. In mine waters microorganisms form a complex microbiological biocoenosis in spite of low pH values and high heavy metal concentrations including high uranium concentrations. The analyses used microbiological methods like confocal laser scanning microscopy and molecular-biological techniques. The interactions of microorganism with fluorescent radioactive heavy metal ions were performed with TRLFS (time resolved laser-induced fluorescence spectroscopy).

  6. Technical and economical aspects of the recycling of plutonium and uranium in Germany

    International Nuclear Information System (INIS)

    The experience and achievements accumulated in the past with recycling of uranium and plutonium are briefly summarized. The proven technology for the recycling of plutonium and uranium is now available for use on an industrial scale and is ready for optimization. The paper discusses demand for plutonium recycling capacity, fabrication of mixed oxide fuel assemblies in Germany, economic aspects of plutonium recycling, impact of plutonium recycling on natural uranium demand, second generation plutonium recycling, and recycling of uranium

  7. Plutonium and americium in fish, shellfish and seaweed in the Irish environment and their contribution to dose

    International Nuclear Information System (INIS)

    Plutonium and americium activity concentrations in fish and shellfish landed in Ireland in the period 1988 to 1997 are presented. Activity concentrations in fish are low and often below detection limits, while those in mussels and oysters sampled on the northeast coast show no significant signs of decline. The estimated doses to hypothetical typical and heavy seafood consumers remain below 1 μSv yr-1 (committed effective dose).Plutonium activity concentrations measured in Fucus vesiculosus around the Irish coastline have not fallen appreciably in the ten year period between 1986 and 1996. Furthermore, the mean 238Pu/239,240Pu ratio of 0.17±0.05 in Fucus vesiculosus from the west coast of Ireland demonstrates the increasing significance of Sellafield-derived plutonium in those waters. (Copyright (c) 1999 Elsevier Science B.V., Amsterdam. All rights reserved.)

  8. Plutonium and americium in Arctic waters, the North Sea and Scottish and Irish coastal zones (in Fucus, Mytilus and Patella)

    International Nuclear Information System (INIS)

    Plutonium and americium have been measured in surface waters of the Greenland and Barents Seas and in the northern North Sea from 1980 through 1984. Measurements in water and biota, Fucus, Mytilus and Patella, were carried out in North-English and Scottish waters in 1982 and Fucus samples were collected from the Irish coast in 1983. Fallout is found to dominate as a source of 239+240Pu north of latitude 650N, while for 238Pu a substantial fraction originates from European nuclear fuel reprocessing facilities. The 238Pu/239+240Pu isotope ratio provides clear evidence of the transport of effluent plutonium from the latter to Spitsbergen waters. Fallout plutonium in Arctic waters has a residence time of the order of several years, while for Pu from Sellafield we estimate mean residence times of 11-15 months in Scottish waters and, tentatively, 1.5-3 y during transport from the North Channel (north of the Irish Sea) to Spitsbergen. 241Am found in Arctic waters probably originates from the decay of fallout 241Pu and, like Pu, tentatively has a residence time of the order of several years. Americium from Sellafield has an estimated mean residence time of 4-6 months in Scottish waters. (author)

  9. Transport of plutonium, americium, and curium from soils into plants by roots

    International Nuclear Information System (INIS)

    For assessing the dose from radionuclides in agricultural products by ingestion it is necessary to know the soil to plant transfer factors. The literature was entirely investigated, in order to judge the size of the soil to plant transfer factors. In total, 92 publications - from 1948 to 1978 -have been evaluated. As result, transfer factors from 10-9 to 10-3 have been found for Plutonium, and from 10-6 to 1 for Americium. For Curium only few data are available in literature. The considerable variation of the measured transfer factors is based on the dependence of these transfer factors from the ion exchange capacity of soils, from the amount of organic materials, from the pH-value, and from the mode of contamination. There are, in any case, contradictory data, although there has been detected a dependence of the transfer factors from these parameters. Chelating agenst increase the transfer factors to approximately 1300. As well, fertilizers have an influence on the size of the transfer factors - however, the relationships have been scarcely investigated. The distribution of actinides within the individual parts of plants has been investigated. The highest concentrations are in the roots; in the plant parts above ground the concentration of actinides decreases considerably. The most inferior transfer factors were measured for the respective seed or fruits. The soil to plant transfer factors of actinides are more dependend on the age of the plants within one growing period. At the beginning of the period, the transfer factor is considerably higher than at the end of this period. With respect to plants with a growing period of several years, correlations are unknown. (orig.)

  10. Carbothermic synthesis of carbides of uranium and plutonium

    International Nuclear Information System (INIS)

    Partial pressures of carbon monoxide, uranium and plutonium over different phase regions relevant to the carbothermic synthesis of carbides of uranium and plutonium are calculated using recent models and thermodynamic data for the compounds in U-C-O and Pu-C-O systems. The experimental parameters for the preparation of uranium carbides and a two step synthesis involving carbothermic reduction of the oxide to the dicarbide followed by hydrogen stripping of carbon to produce uranium monocarbide are discussed. (author). 31 refs., 9 figs., 6 tabs

  11. Recovery and purification of uranium-234 from aged plutonium-238

    International Nuclear Information System (INIS)

    The current production methods used to recover and purify uranium-234 from aged plutonium-238 at Mound Laboratory are presented. The three chemical separation steps are described in detail. In the initial separation step, the bulk of the plutonium is precipitated as the oxalate. Successively lower levels of plutonium are achieved by anion exchange in nitrate media and by anion exchange in chloride media. The procedures used to characterize and analyze the final U3O8 are given

  12. Diffusion in the uranium - plutonium system and self-diffusion of plutonium in epsilon phase

    International Nuclear Information System (INIS)

    A survey of uranium-plutonium phase diagram leads to confirm anglo-saxon results about the plutonium solubility in α uranium (15 per cent at 565 C) and the uranium one in ζ phase (74 per cent at 565 C). Interdiffusion coefficients, for concentration lower than 15 per cent had been determined in a temperature range from 410 C to 640 C. They vary between 0.2 and 6 1012 cm2 s-1, and the activation energy between 13 and 20 kcal/mole. Grain boundary, diffusion of plutonium in a uranium had been pointed out by micrography, X-ray microanalysis and α autoradiography. Self-diffusion of plutonium in ε phase (bcc) obeys Arrhenius law: D = 2. 10-2 exp -(18500)/RT. But this activation energy does not follow empirical laws generally accepted for other metals. It has analogies with 'anomalous' bcc metals (βZr, βTi, βHf, Uγ). (author)

  13. Americium 241 in vegetation of natural biocenoses and agrocenoses on Belarus territories contaminated with Chernobyl fall-out

    International Nuclear Information System (INIS)

    As a result of beta-decay of plutonium 241 the content of americium 241 increases progressively in soils, contaminated with Chernobyl trans uranium elements. Americium 241 displayed higher (0,5 - 1,5-fold) biological mobility than isotopes of plutonium 239, 240. Activity of americium 241 in surface phyto mass of wild and cultural plants varies from 0,04 to 5,9 Bq/kg of dry weight. Americium 241 contribution to the total trans uranium elements contamination of plants made up 60 - 80% in 1996 - 1998. Investigation of trials from the areas adjacent to the 30-km zone showed that mobility of americium 241 and plutonium was 5 - 15 times as high as in the zone

  14. Recovery, purification and concentration of plutonium and americium from the aqueous wastes discharged in the reprocessing process studies

    International Nuclear Information System (INIS)

    For recovering and purifying plutonium and americium from the aqueous wastes occurring in the process studies on reprocessing, a standard procedure has been established for use in the laboratory works, through the preliminary tests of the precipitation as hydroxides and the anion exchange in nitrate media. The procedure was proven in the treatment of actual wastes, of which the results were contributed to determine the process conditions in the plutonium purification and product concentration of the JAERI Reprocessing Test Plant. The preliminary tests also include washing of U and Am recovery from the anion-exchanger in nitrate media, direct ion-exchange recovery of Pu from the TBP phase and elution of Am from the cation-exchanger. (auth.)

  15. Fluorination of uranium and plutonium dioxides in flame tower reactors

    International Nuclear Information System (INIS)

    Fluorination of uranium and plutonium dioxides by fluorine is discussed for the flame tower reactor, which is one of possible reactors is FLUOREX process, a hybrid reprocessing combining a fluoride volatility process and a solvent extraction. A reaction model for a shrinking particle with an un-reacted shrinking core is applied to the fluorination of plutonium dioxide and the reaction rate constants are determined in low temperature range around 400degC using the reported kinetic data [Iwasaki et al., J. Nucl. Sci. Technol., 11, 403 (1974)]. The determined rate equations are used to analyze the fluorination of uranium and plutonium dioxides in the high temperature range (800-1,200degC), where the flame tower reactor is operated. For uranium dioxides the diffusion of fluorine to the particle controls the overall fluorination reaction, and the model can predict the recovery of uranium to the product of uranium hexafluoride. For plutonium dioxides, however, the recoveries of plutonium calculated by the presented model do not agree with the experimental recoveries. It is suggested that the reaction mechanism for the fluorination of plutonium dioxide in the high temperature range is different from that in the low temperature range. (author)

  16. Osteosarcoma induction by plutonium-239, americium-241 and neptunium-237 : the problem of deriving risk estimates for man

    International Nuclear Information System (INIS)

    Spontaneous bone cancer (osteosarcoma) represents only about 0.3% of all human cancers, but is well known to be inducible in humans by internal contamination with radium-226 and radium-224. plutonium-239, americium-241 and neptunium-237 form, or will form, the principal long-lived alpha particle emitting components of high activity waste and burnt-up nuclear fuel elements. These three nuclides deposit extensively in human bone and although, fortunately, no case of a human osteosarcoma induced by any of these nuclides is known, evidence from animal studies suggests that all three are more effective than radium-226 in inducing osteosarcoma. The assumption that the ratio of the risk factors, the number of osteosarcoma expected per 10000 person/animal Gy, for radium-226 and any other bone-seeking alpha-emitter will be independent of animal species has formed the basis of all the important studies of the radiotoxicity of actinide nuclides in experimental animals. The aim of this communication is to review the risk factors which may be calculated from the various animal studies carried out over the last thirty years with plutonium-237, americium-241 and neptunium-237 and to consider the problems which may arise in extrapolating these risk factors to homo sapiens

  17. Modified Purex process for the separation and recovery of plutonium--uranium residues

    Energy Technology Data Exchange (ETDEWEB)

    Navratil, J.D.; Leebl, R.G.

    1978-07-08

    A modified (one-cycle) Purex process has been developed for the separation and recovery of plutonium and uranium from mixed actinide residues. The process utilizes 30 vol % tributyl phosphate--dodecane to extract uranium from a 5M nitric acid-plutonium (III)-uranium(VI) feed. After uranium extraction, plutonium in the aqueous feed solution is purified by anion exchange technology. Uranium in the organic is scrubbed and stripped to effectively purify the uranium so that it contains <5,000 ppM plutonium. The process has been used successfully to separate residues consisting of plutonium and uranium oxide.

  18. Characteristics of plutonium and americium contamination at the former U.K. atomic weapons test ranges at Maralinga and Emu

    Energy Technology Data Exchange (ETDEWEB)

    Burns, P.A.; Cooper, M.B.; Lokan, K.H.; Wilks, M.J.; Williams, G.A. [Australian Radiation Lab., Melbourne, VIC (Australia)

    1995-11-01

    Physico-chemical studies on environmental plutonium are described, which provide data integral to an assessment of dose for the inhalation of artificial actinides by Australian Aborigines living a semi-traditional lifestyle at Maralinga and Emu, sites of U.K. atomic weapons tests between 1953 and 1963. The most significant area, from a radiological perspective, is the area contaminated by plutonium in a series of ``one point`` safety trials in which large quantities of plutonium were dispersed explosively at a location known as Taranaki. The activity distribution of plutonium and americium with particle size is quite different from the mass distribution, as a considerably higher proportion of the activity is contained in the finer (inhalable) fraction than of the mass. Except in areas which were disturbed through ploughing during a cleanup in 1967, most the activity remains in the top 1 cm of the surface. Much of the activity is in particulate form, even at distances > 20 km from the firing sites, and discrete particles have been located even at distances beyond 100 km. Data are presented which permit the assessment of annual committed doses through the inhalation pathway, for Aborigines living a semi-traditional lifestyle in the areas affected by the Taranaki firings in particular. (author).

  19. Characteristics of plutonium and americium contamination at the former U.K. atomic weapons test ranges at Maralinga and Emu

    International Nuclear Information System (INIS)

    Physico-chemical studies on environmental plutonium are described, which provide data integral to an assessment of dose for the inhalation of artificial actinides by Australian Aborigines living a semi-traditional lifestyle at Maralinga and Emu, sites of U.K. atomic weapons tests between 1953 and 1963. The most significant area, from a radiological perspective, is the area contaminated by plutonium in a series of ''one point'' safety trials in which large quantities of plutonium were dispersed explosively at a location known as Taranaki. The activity distribution of plutonium and americium with particle size is quite different from the mass distribution, as a considerably higher proportion of the activity is contained in the finer (inhalable) fraction than of the mass. Except in areas which were disturbed through ploughing during a cleanup in 1967, most the activity remains in the top 1 cm of the surface. Much of the activity is in particulate form, even at distances > 20 km from the firing sites, and discrete particles have been located even at distances beyond 100 km. Data are presented which permit the assessment of annual committed doses through the inhalation pathway, for Aborigines living a semi-traditional lifestyle in the areas affected by the Taranaki firings in particular. (author)

  20. Preconcentration of low levels of americium and plutonium from waste waters by synthetic water-soluble metal-binding polymers with ultrafiltration

    International Nuclear Information System (INIS)

    A preconcentration approach to assist in the measurement of low levels of americium and plutonium in waste waters has been developed based on the concept of using water-soluble metal-binding polymers in combination with ultrafiltration. The method has been optimized to give over 90% recovery and accountability from actual waste water. (author)

  1. Determination of Uranium and Plutonium Concentration in 1AF by Isotopic Dilution Mass Spectrometry Methods

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>It is important data to measure uranium and plutonium concentration for the reprocessing plant control analysis. The determination of uranium and plutonium concentration in 1AF by isotopic dilution mass

  2. Study of reactions between uranium-plutonium mixed oxide and uranium nitride and between uranium oxide and uranium nitride

    International Nuclear Information System (INIS)

    A new type of combustible elements which is a mixture of uranium nitride and uranium-plutonium oxide could be used for Quick Neutrons Reactors. Three different studies have been made on the one hand on the reactions between uranium nitride (UN) and uranium-plutonium mixed oxide (U,Pu)O2, on the other hand on these between UN and uranium oxide UO2. They show a sizeable reaction between nitride and oxide for the studied temperatures range (1573 K to 1973 K). This reaction forms a oxynitride compound, MOx Ny with M=U or M=(U,Pu), whose crystalline structure is similar to oxide's. Solubility of nitride in both oxides is studied, as the reaction kinetics. (TEC). 32 refs., 48 figs., 22 tabs

  3. Ultratrace analysis of uranium and plutonium by mass spectrometry

    International Nuclear Information System (INIS)

    Full text: Uranium and plutonium have traditionally been analyzed using alpha energy spectrometry. Both isotopic compositions and elemental abundances can be characterized on samples containing microgram to milligram quantities of uranium and nanogram to microgram quantities of plutonium. In the past ten years or so, considerable interest has developed in measuring nanograms quantities of uranium and sub-picogram quantities of plutonium in environmental samples. Such measurements require high sensitivity and as a consequence, sensitive mass spectrometric-based methods have been developed. Thus, the analysis of uranium and plutonium have gone from counting decays to counting atoms, with considerable increases in both sensitivity and precision for isotopic measurements. At the Pacific Northwest National Laboratory (PNNL), we have developed highly sensitive methods to analyze uranium and plutonium in environmental samples. The development of an ultratrace analysis capability for measuring uranium and plutonium has arisen from a need to detect and characterize environmental samples for signatures associated with nuclear industry processes. Our most sensitive well-developed methodologies employ thermal ionization mass spectrometry (TIMS), however, recent advances in inductively coupled plasma mass spectrometry (ICP-MS) have shown considerable promise for use in detecting uranium and plutonium at ultratrace levels. The work at PNNL has included the development of both chemical separation and purification techniques, as well as the development of mass spectrometric instrumentation and techniques. At the heart of our methodology for TIMS analysis is a procedure that utilizes 100-microliter-volumes of analyte for chemical processing to purify, separate, and load actinide elements into resin beads for subsequent mass spectrometric analysis. The resin bead technique has been combined with a thorough knowledge of the physicochemistry of thermal ion emission to achieve

  4. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    International Nuclear Information System (INIS)

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

  5. Phenomenology of uranium-plutonium homogenization in nuclear fuels

    International Nuclear Information System (INIS)

    The uranium and plutonium cations distribution in mixed oxide fuels (U1-y Puy)O2 with y ≤ 0.1 has been studied in laboratory with industrial fabrication methods. Our experiences has showed a slow cations migration. In the substoichiometry (UPu)O2-x the diffusion is in connection with the plutonium valence which is an indicator of the oxidoreduction state of the crystal lattice. The plutonium valence is in connection with the oxygen ion deficit in order to compensate the electrical charge. The oxygen ratio of the solid depends of the oxygen partial pressure prevailing at the time of product elaboration but it can be modified by impurities. These impurities permit to increase or decrease the fuel characteristics and performances. An homogeneity analysis methodology is proposed, its objective is to classify the mixed oxide fuels according to the uranium and plutonium ions distribution

  6. Standard specification for sintered (Uranium-Plutonium) dioxide pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This specification covers finished sintered and ground (uranium-plutonium) dioxide pellets for use in thermal reactors. It applies to uranium-plutonium dioxide pellets containing plutonium additions up to 15 % weight. This specification may not completely cover the requirements for pellets fabricated from weapons-derived plutonium. 1.2 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all applicable international, federal, state, and local regulations pertaining to possessing, processing, shipping, or using source or special nuclear material. Examples of U.S. government documents are Code of Federal Regulations Title 10, Part 50Domestic Licensing of Production and Utilization Facilities; Code of Federal Regulations Title 10, Part 71Packaging and Transportation of Radioactive Material; and Code of Federal Regulations Tit...

  7. HPAT: A nondestructive analysis technique for plutonium and uranium solutions

    International Nuclear Information System (INIS)

    Two experimental approaches for the nondestructive characterization of mixed solutions of plutonium and uranium, developed at BNEA - C.R.E. Casaccia, with the goal of measuring low plutonium concentration (<50 g/l) even in presence of high uranium content, are described in the following. Both methods are referred to as HPAT (Hybrid Passive-Active Technique) since they rely on the measurement of plutonium spontaneous emission in the LX-rays energy region as well as the transmission of KX photons from the fluorescence induced by a radioisotopic source on a suitable target. Experimental campaigns for the characterization of both techniques have been carried out at EUREX Plant Laboratories (C.R.E. Saluggia) and at Plutonium Plant Laboratories (C.R.E. Casaccia). Experimental results and theoretical value of the errors are reported. (author)

  8. Separation Method of Uranium and Plutonium From Large Amount of Neptunium

    Institute of Scientific and Technical Information of China (English)

    SU; Yu-lan; JIN; Hua; YING; Zhe-cong; ZHAO; Sheng-yang

    2013-01-01

    Uranium and plutonium are limited strictly in the neptunium product.To eliminate the influence of neptunium matrix on determination of uranium and plutonium,a new separation method of uranium and plutonium from large amount of neptunium by TEVA column has been developed,which is illustrated in Fig.1.

  9. Pyrochemical reduction of uranium dioxide and plutonium dioxide by lithium metal

    International Nuclear Information System (INIS)

    The lithium reduction process has been developed to apply a pyrochemical recycle process for oxide fuels. This process uses lithium metal as a reductant to convert oxides of actinide elements to metal. Lithium oxide generated in the reduction would be dissolved in a molten lithium chloride bath to enhance reduction. In this work, the solubility of Li2O in LiCl was measured to be 8.8 wt% at 650 deg. C. Uranium dioxide was reduced by Li with no intermediate products and formed porous metal. Plutonium dioxide including 3% of americium dioxide was also reduced and formed molten metal. Reduction of PuO2 to metal also occurred even when the concentration of lithium oxide was just under saturation. This result indicates that the reduction proceeds more easily than the prediction based on the Gibbs free energy of formation. Americium dioxide was also reduced at 1.8 wt% lithium oxide, but was hardly reduced at 8.8 wt%

  10. Extraction and chromatographic separation and concentration of plutonium and americium from natural matrices. Author-review of dissertation submitted for fulfillment of the scientific degree 'Philosophiae doctor' (PhD.)

    International Nuclear Information System (INIS)

    We followed the optimization of separation progress of americium (241Am) from environmental samples - soil from surroundings nuclear power plant Jaslovske Bohunice, in our work. Selection and optimization of separation progresses had to verify the condition for preparation of samples on spectral measurement with coprecipitation of americium or plutonium with NdF3 (undesirable presence of calcium, magnesium, lanthanides) and condition of spectral purity (spectral overlapping 228Th, 238Pu, 241Am and 222Rn of energy). Very important step was the realization of existing goal and learn suitable isolation techniques of plutonium. We are choosing technique separation of plutonium base upon amine liquid extraction, for a digest consider qualitative quantitative factor of separation. Extraction reagent has been Aliquat-336, which extracts nitric complex of plutonium [Pu(NO3)62-] from 7-8 M solution HNO3. Use method separate off quantitative the plutonium, thorium and uranium from americium. Background sample formed the sample of soil from surroundings Velke Kostolany. Real samples were sampling from surroundings of pollute river Dudvah. Average value mass activities of 239,240Pu in the background sample had value 0.28 ± 0.10 Bq · kg-1. Value mass activities of 239,240Pu in sample from surroundings river Dudvah were in the range (0,6 - 39.4) Bq · kg-1. Methodical side separation of americium we step by step by using ion exchange methods, liquid extraction with extraction reagent TOPO, or combination of them and extraction chromatography with TOPO. We find out: (a) on exchange procedure are suitable on obtainable basis extract tracer of radionuclide, also is very up to time. Optimal method was indicate techniques using the formation of rhodanide complex of americium, with following adsorption on stark acidity anionic exchanger (lanthanides were non-absorbing); (b) t liquid extraction formed emulsion, the third phase on the interface of phases. If we treat the molar of

  11. METHOD FOR PREPARING URANIUM MONOCARBIDE-PLUTONIUM MONOCARBIDE SOLID SOLUTION

    Science.gov (United States)

    Ogard, A.E.; Leary, J.A.; Maraman, W.J.

    1963-03-19

    A method is given for preparing solid solutions of uranium monocarbide- plutonium monocarbide. In this method, the powder form of uranium dioxide, plutonium dioxide, and graphite are mixed in a ratio determined by the equation: xUO/sub 2/ + yPuO/sub 2/ + (2+z)C yields UxPu/sub y/C/sub z/ +2CO, where x + y equ al 1.0 and z is greater than 0.9 but less than 1.0. The resulting mixture is compacted and heated in a vacuum at a temperature of 1850 deg C. (AEC)

  12. Determination of oxygen in mixed uranium-plutonium carbide fuels

    International Nuclear Information System (INIS)

    Determination of oxygen in mixed uranium-plutonium carbide fuels is made by inert gas fusion-coulometry. To minimize oxygen contamination during sample preparation, the sample is crushed, weighted and sealed air-tight in a platinum capsule in an argon gas atmosphere glove box. The true oxygen content is estimated by subtracting the oxygen contamination from the oxygen determined. Routine analysis of 32 samples of mixed uranium-plutonium carbides is performed with a coefficient of variation of 1.6%. (author)

  13. The bone volume effect on the dosimetry of plutonium-239 and americium-241 in the skeleton of man and baboon

    International Nuclear Information System (INIS)

    Studies were undertaken using bone removed from young adult baboons, which had been contaminated with plutonium-239 at various times prior to sacrifice, and human bone from adult male (USTR Case 246), who had received an internal deposition of americium-241 as a result of a glove-box explosion 11 years prior to his death. The baboon bone was supplied by the CEA, France, and the human bone by the United States Transuranium registry. The bone samples, examined by qualitative and quantitative autoradiography with CR 39 detectors, demonstrated the rapid redistribution of bone surface-seeking radionuclides in younger primates due to growth and the slower, bone turnover driven redistribution in the adult human bone. In both species, primary and secondary surface deposits of radionuclide remained conspicious despite bone activity; true volumization of radionuclide was seldom seen. The dosimetric implications of these findings are discussed. (author) 21 refs.; 6 figs.; 4 tabs

  14. An analysis of experiments on graphite moderated assemblies fuelled with uranium, plutonium and uranium-plutonium metal

    International Nuclear Information System (INIS)

    Several sets of experiments have been studied as a justification for the nuclear data and computational models, particularly the TRACER and MINX computer codes, currently in use at AEE Winfrith for the calculation of graphite moderated reactors. Buckling measurements are compared to theory for systems fuelled with plutonium/aluminium and uranium 235/aluminium spikes, and with uranium and uranium-plutonium metal rods. In some cases reaction-rates and temperature coefficients are also available, and comparisons with these are reported. Agreement between theory and experiment is generally good. The study does, however, point to some defects in the models used, and suggests that the value of eta used for plutonium (2.088 at 2200 m/s) is slightly too low, as would be expected from the latest differential measurements. (author)

  15. Plutonium well logging with the photoneutron uranium exploration system

    International Nuclear Information System (INIS)

    The Los Alamos National Laboratory prototype photoneutron uranium exploration system was recently demonstrated at the Hanford site near Richland, Washington, for Rockwell-Hanford Operations (Rockwell). The demonstration determined the field performance capabilities of the uranium exploration system for in situ, downhole measurements of transuranic waste concentrations. The uranium exploration system is indeed capable of detecting plutonium in the test wells at the waste sites investigated. The excellent signal-to-background ratio (15:1 in the worst case) of the system made positive plutonium determinations possible despite neutron backgrounds caused by spontaneous fission and (α,n) emitters. We present all the data collected from seven test wells and guidance for interpreting the data relative to the known uranium ore calibration of the system. The demonstration indicated no operational difficulties in the waste site environment, and routine use by Rockwell personnel appears practical

  16. On-line monitoring of plutonium in mixed uranium-plutonium solutions

    International Nuclear Information System (INIS)

    The measurement of the total and isotopic plutonium concentrations in mixed uranium-plutonium solutions blended with highly radioactive fission product nuclides and other radionuclides (e.g., Cs-137 and Co-60) has been investigated at the Barnwell Nuclear Fuel Plant (BNFP). An on-line total and isotopic plutonium monitoring system is being tested for its ability to assay the plutonium abundances in solutions as might be found in the process streams of a light water reactor (LWR) spent fuel processing plant. The monitoring system is fully automated and designed to be maintained remotely. It is capable of near real-time inventory of plutonium in process streams and provides the basis for on-line computerized accounting of special nuclear materials

  17. Determination of uranium and plutonium in high active solutions by extractive spectrophotometry

    International Nuclear Information System (INIS)

    Plutonium and uranium was extracted from nitric acid into trioctyl phosphine oxide in xylene. The TOPO layer was analysed by spectrophotometry. Thoron was used as the chromogenic agent for plutonium. Pyridyl azoresorcinol was used as chromogenic agent for uranium. The molar absorption coefficient for uranium and plutonium was found to be 19000 and 19264 liter/mole-cm, respectively. The correlation coefficient for plutonium and uranium was found to be 0.9994. The relative standard deviation for the determination of plutonium and uranium was found to be 0.96% and 1.4%, respectively. (author)

  18. Safe handling of kilogram amounts of fuel-grade plutonium and of gram amounts of plutonium-238, americium-241 and curium-244

    International Nuclear Information System (INIS)

    During the past 10 years about 600 glove-boxes have been installed at the Institute for Transuranium Elements at Karlsruhe. About 80% of these glove-boxes have been designed and equipped for handling 100-g to 1-kg amounts of 239Pu containing 8-12% 240Pu (low-exposure plutonium). A small proportion of the glove-boxes is equipped with additional shielding in the form of lead sheet or lead glass for work with recycled plutonium. In these glove-boxes gram-amounts of 241Am have also been handled for preparation of Al-Am targets using tongs and additional shielding inside the glove-boxes themselves. Water- and lead-shielded glove-boxes equipped with telemanipulators have been installed for routine work with gram-amounts of 241Am, 243Am and 244Cm. A prediction of the expected radiation dose for the personnel is difficult and only valid for a preparation procedure with well-defined preparation steps, owing to the fact that gamma dose-rates depend strongly upon proximity and source seize. Gamma radiation dose measurements during non-routine work for 241Am target preparation showed that handling of gram amounts leads to a rather high irradiation dose for the personnel, despite lead or steel glove-box shielding and shielding within the glove-boxes. A direct glove-hand to americium contact must be avoided. For all glove-handling of materials with gamma radiation an irradiation control of the forearms of the personnel by, for example, thermoluminescence dosimeters is necessary. Routine handling of americium and curium should be executed with master-slave equipment behind neutron and gamma shielding. (author)

  19. Resuspension of uranium-plutonium oxide particles from burning Plexiglas

    International Nuclear Information System (INIS)

    Nuclear fuel materials such as Uranium-Plutonium oxide must be handled remotely in gloveboxes because of their radiotoxicity. These gloveboxes are frequently constructed largely of combustible Plexiglas sheet. To estimate the potential airborne spread of radioactive contamination in the event of a glovebox fire, the resuspension of particles from burning Plexiglas was investigated. (author)

  20. Polonium, uranium and plutonium in the southern Baltic ecosystem

    International Nuclear Information System (INIS)

    This paper presents the results of the measurement of polonium, uranium and plutonium alpha radionuclides in seawater and biota of the southern Baltic ecosystem as well as the recognition of their accumulation processes in the trophic chain. Investigation of the polonium 210Po and plutonium 239+240Pu concentrations in Baltic biota revealed that these radionuclides are strongly accumulated by some species. Mean values of the bioconcentration factor (BCF) fell within the range 9x102 to 3.7x104. The Baltic Sea algae, benthic animals and fish concentrated uranium radioisotopes only to a small extent and mean BCF values for this element range from 1 to 55, which is several orders of magnitude lower than that for polonium and plutonium. Moreover, it was found that Baltic fish constitute an important source of polonium 210Po for humans. (author)

  1. Distribution of plutonium, americium, and several rare earth fission product elements between liquid cadmium and LiCl-KCl eutectic

    International Nuclear Information System (INIS)

    Separation factors were measured that describe the partition between molten cadmium and molten LiCl-KCl eutectic of plutonium, americium, praseodymium, neodymium, cerium, lanthanum, gadolinium, dysprosium, and yttrium. The temperature range was 753-788 K, and the range of concentrations was that allowed by the sensitivity of the chemical analysis methods. Mean separation factors were derived for Am-Pu, Nd-Am, Nd-Pu, Nd-Pr, Gd-La, Dy-La, La-Ce, La-Nd, Y-La, and Y-Nd. Where previously published data were available, agreement was good. For convenience, the following series of separation factors relative to plutonium was derived by combining the measured separation factors: Pu, 1.00 (basis); Am, 1.54; Pr, 22.0; Nd, 23.4; Ce, 26; La, 70; Gd, 77; Dy, 270; Y, 3000. These data are used in calculating the distribution of the actinide and rare earth elements in the prochemical reprocessing of spent fuel from the Integral Fast Reactor. (orig.)

  2. Determination by gamma-ray spectrometry of the plutonium and americium content of the Pu/Am separation scraps. Application to molten salts; Determination par spectrometrie gamma de la teneur en plutonium et en americium de produits issus de separation Pu/Am. Application aux bains de sels

    Energy Technology Data Exchange (ETDEWEB)

    Godot, A. [CEA Valduc, Dept. de Traitement des Materiaux Nucleaires, 21 - Is-sur-Tille (France); Perot, B. [CEA Cadarache, Dept. de Technologie Nucleaire, Service de Modelisation des Transferts et Mesures Nucleaires, 13 - Saint-Paul-lez-Durance (France)

    2005-07-01

    Within the framework of plutonium recycling operations in CEA Valduc (France), americium is extracted from molten plutonium metal into a molten salt during an electrolysis process. The scraps (spent salt, cathode, and crucible) contain extracted americium and a part of plutonium. Nuclear material management requires a very accurate determination of the plutonium content. Gamma-ray spectroscopy is performed on Molten Salt Extraction (MSE) scraps located inside the glove box, in order to assess the plutonium and americium contents. The measurement accuracy is influenced by the device geometry, nuclear instrumentation, screens located between the sample and the detector, counting statistics and matrix attenuation, self-absorption within the spent salt being very important. The purpose of this study is to validate the 'infinite energy extrapolation' method employed to correct for self-attenuation, and to detect any potential bias. We present a numerical study performed with the MCNP computer code to identify the most influential parameters and some suggestions to improve the measurement accuracy. A final uncertainty of approximately 40% is achieved on the plutonium mass. (authors)

  3. Plutonium, americium and radiocaesium in the marine environment close to the Vandellos I nuclear power plant before decommissioning

    International Nuclear Information System (INIS)

    The Vandellos nuclear power plant (NPP), releasing low-level radioactive liquid waste to the Mediterranean Sea, is the first to be decommissioned in Spain, after an incident which occurred in 1989. The presence, distribution and uptake of various artificial radionuclides (radiocaesium, plutonium and americium) in the environment close to the plant were studied in seawater, bottom sediments and biota, including Posidonia oceanica, fish, crustaceans and molluscs. Seawater, sediments and Posidonia oceanica showed enhanced levels in the close vicinity of the NPP, although the effect was restricted to its near environment. Maximum concentrations in seawater were 11.6±0.5 Bq m-3 and 16.9±1.2 mBq m-3 for 137Cs and 239,240Pu, respectively. When sediment concentrations were normalized to excess 210Pb, they showed both the short-distance transport of artificial radionuclides from the Vandellos plant and the long-distance transport of 137Cs from the Asco NPP. Posidonia oceanica showed the presence of various gamma-emitters attributed to the impact of the Chernobyl accident, on which the effect of the NPP was superimposed. Seawater, sediment and Posidonia oceanica collected near the plant also showed an enhancement of the plutonium isotopic ratio above the fallout value. The uptake of these radionuclides by marine organisms was detectable but limited. Pelagic fish showed relatively higher 137Cs concentrations and only in the case of demersal fish was the plutonium isotopic ratio increased. The reported levels constitute a set of baseline values against which the impact of the decommissioning operations of the Vandellos I NPP can be studied

  4. A new method for the determination of plutonium and americium using high pressure microwave digestion and alpha-spectrometry or ICP-SMS

    International Nuclear Information System (INIS)

    Plutonium and americium are radionuclides particularly difficult to measure in environmental samples because they are a-emitters and therefore necessitate a careful separation before any measurement, either using radiometric methods or ICP-SMS. Recent developments in extraction chromatography resins such as EichromR TRU and TEVA have resolved many of the analytical problems but drawbacks such as low recovery and spectral interferences still occasionally occur. Here, we report on the use of the new EichromR DGA resin in association with TEVA resin and high pressure microwave acid leaching for the sequential determination of plutonium and americium in environmental samples. The method results in average recoveries of 83 ± 15% for plutonium and 73 ± 22% for americium (n = 60), and a less than 10% deviation from reference values of four IAEA reference materials and three samples from intercomparisons exercises. The method is also suitable for measuring 239Pu in water samples at the μBq/l level, if ICP-SMS is used for the measurement. (author)

  5. Contribution to the prediction of americium, plutonium and neptunium behaviour in the geosphere: chemical data

    International Nuclear Information System (INIS)

    An exhaustive bibliographic review on hydrolysis of americium gives the stability constants, at zero ionic strength. No evidence of Am(OH)4- formation was found by solubility studies up to pH 2 (CO3)3 characterised by its X-ray diffraction pattern is studied at a high ionic strength. All the published results on Am in carbonate media are reinterpreted using these stability constants (Am-OH-CO3 complexes are not needed). No evidence of Am(CO3)45- formation was found by spectrophotometry up to 3M. Literature results are used to determine the formal redox potentials at pH = 9.4 and to calculate the formation constants, at zero ionic strength. The formation of complexes between americium and humic materials (purified fulvic and humic acids) has been studied by a spectrophotometric technique. The results are interpreted by the formation of a 1:1 complexe. Solubility of the solid PuO2(CO3) is measured in bicarbonate media at high ionic strength, to obtain the solubility product and formation constants of the PuO2(CO3)i2-2i complexes

  6. A solvent proceed for the extraction of the irradiate uranium and plutonium in the reactor core

    International Nuclear Information System (INIS)

    Description of the conditions of plutonium, fission products and of uranium separation by selective extraction of the nitrates by organic solvent, containing a simultaneous extraction of plutonium and uranium, followed by a plutonium re-extraction after reduction, and an uranium re-extraction. The rates of decontamination being insufficient in this first stage, we also describes the processes of decontamination permitting separately to get the rates wanted for uranium and plutonium. Finally, we describes the beginning of the operation that consists in a nitric dissolution of the active uranium while capturing the products of gaseous fission, as well as the final concentration of the products of fission in a concentrated solution. (authors)

  7. An analysis of the impact of having uranium dioxide mixed in with plutonium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    MARUSICH, R.M.

    1998-10-21

    An assessment was performed to show the impact on airborne release fraction, respirable fraction, dose conversion factor and dose consequences of postulated accidents at the Plutonium Finishing Plant involving uranium dioxide rather than plutonium dioxide.

  8. An analysis of the impact of having uranium dioxide mixed in with plutonium dioxide

    International Nuclear Information System (INIS)

    An assessment was performed to show the impact on airborne release fraction, respirable fraction, dose conversion factor and dose consequences of postulated accidents at the Plutonium Finishing Plant involving uranium dioxide rather than plutonium dioxide

  9. The plutonium-oxygen and uranium-plutonium-oxygen systems: A thermochemical assessment

    International Nuclear Information System (INIS)

    The report of a panel of experts convened by the IAEA in Vienna in March 1964. It reviews the structural and thermodynamic data for the Pu-O and U-Pu-O systems and presents the conclusions of the panel. The report gives information on preparation, phase diagrams, thermodynamic and vaporization behaviour of plutonium oxides, uranium-plutonium oxides and PuO2-MeOx (Me=Be, Mg, Al, Si, W, Th, Eu, Zr, Ce) systems. 167 refs, 27 figs, 17 tabs

  10. Evaporation behaviour of the ternary uranium plutonium carbides

    International Nuclear Information System (INIS)

    The evaporation behaviour of uranium plutonium carbides (Usub(0.80)Psub(0.20)Csub(1+-x) was studied by a combined application of mass spectrometry, using the uranium isotope U-233, and the Knudsen effusion target collection technique in the temperature range from 15000C to the liquids temperature measured at 24580C and the composition range from C/M = 0.95 to 1.4. High temperature compatibility tests were made with W-cells, carburized Ta and TaC-liners up to 25000C. The influence of oxygen and nitrogen impurities on vapour pressure, and composition changes in continued evaporation of the the mixed carbides were investigated. The effects of plutonium depletion and segregation were studied. (Auth.)

  11. Study of the reaction of uranium and plutonium with bone char

    International Nuclear Information System (INIS)

    A study of the reaction of plutonium with a commercial bone char indicates that this bone char has a high capacity for removing plutonium from aqueous wastes. The adsorption of plutonium by bone char is pH dependent, and for plutonium(IV) polymer appears to be maximized near pH 7.3 for plutonium concentrations typical of some waste streams. Adsorption is affected by dissolved salts, especially calcium and phosphate salts. Freundlich isotherms representing the adsorption of uranium and plutonium have been prepared. The low potential imposed upon aqueous solutions by commercial bone char is adequate for reduction of hexavalent plutonium to a lower plutonium oxidation state

  12. Uranium and plutonium in marine sediments

    International Nuclear Information System (INIS)

    The marine sediments contain uranium concentrations that are considered normal, since the seawater contains dissolved natural uranium that is deposited in the bed sea in form of sediments by physical-chemistry and bio-genetics processes. Since the natural uranium is constituted of several isotopes, the analysis of the isotopic relationship 234U/238U are an indicator of the oceanic activity that goes accumulating slowly leaving a historical registration of the marine events through the profile of the marine soil. But the uranium is not the only radioelement present in the marine sediments. In the most superficial strata the presence of the 239+140Pu has been detected that it is an alpha emitter and that recently it has been detected with more frequency in some coasts of the world. The Mexican coast has not been the exception to this phenomenon and in this work the presence of 239-140Pu is shown in the more superficial layers of an exploring coming from the Gulf of Tehuantepec. (Author)

  13. Evaluation of synthetic water-soluble metal-binding polymers with ultrafiltration for selective concentration of americium and plutonium

    International Nuclear Information System (INIS)

    Routine counting methods and ICP-MS are unable to directly measure the new US Department of Energy (DOE) regulatory level for discharge waters containing alpha-emitting radionuclides of 30 pCi/L total alpha or the 0.05 pCi/L regulatory level for Pu or Am activity required for surface waters at the Rocky Flats site by the State of Colorado. This inability indicates the need to develop rapid, reliable, and robust analytical techniques for measuring actinide metal ions, particularly americium and plutonium. Selective separation or preconcentration techniques would aid in this effort. Water-soluble metal-binding polymers in combination with ultrafiltration are shown to be an effective method for selectively removing dilute actinide ions from acidic solutions of high ionic strength. The actinide-binding properties of commercially available water-soluble polymers and several polymers which have been reported in the literature were evaluated. The functional groups incorporated in the polymers were pyrrolidone, amine, oxime, and carboxylic, phosphonic, or sulfonic acid. The polymer containing phosphonic acid groups gave the best results with high distribution coefficients and concentration factors for 241Am(III) and 238Pu(III)/(IV) at pH 4 to 6 and ionic strengths of 0.1 to 4

  14. Migration of the fission products strontium, technetium, iodine, cesium, and the actinides neptunium, plutonium, americium in granitic rock

    International Nuclear Information System (INIS)

    Rock samples were taken from drilling cores in granitic and granodioritic rock, and small (2x2x2 cm) rock tablets from the drilling cores were exposed to a groundwater solution containing one of the studied elements at race levels. The concentration of the element versus penetration depth in the rock tablet was measured radiometrically. The sorption on the mineral faces and the migration into the rock was studied, by an autoradiographic technique. The cationic fission products strontium and cesium had apparent diffusivities of 10-13-10-14 m2/s. They migrate mainly in fissures or filled fractures containing e.g., calcite, epidote or chlorite or in veins with hgih capacity minerals (e.g. biotite). The anionic fission products iodine and technetium had apparent diffusivities of about 10-14 m2/s. These species migrate along mineral boundaries and in open fractures and to a minor extent in high capacity mineral veins. The migration of the actinides neptunium, plutonium and americium is very slow (in the mm-range after 2-3 years contact time). The apparent diffusivities were about 10-15 m2/s. The actinide migration into the rock was largely confined to fissures. (orig./HP)

  15. Measured solubilities and speciations of neptunium, plutonium, and americium in a typical groundwater (J-13) from the Yucca Mountain region

    International Nuclear Information System (INIS)

    Solubility and speciation data are important in understanding aqueous radionuclide transport through the geosphere. They define the source term for transport retardation processes such as sorption and colloid formation. Solubility and speciation data are useful in verifying the validity of geochemical codes that are part of predictive transport models. Results are presented from solubility and speciation experiments of 237NpO2+, 239Pu4+, 241Am3+/Nd3+, and 243Am3+ in J-13 groundwater (from the Yucca Mountain region, Nevada, which is being investigated as a potential high-level nuclear waste disposal site) at three different temperatures (25 degree, 60 degree, and 90 degree C) and pH values (5.9, 7.0, and 8.5). The solubility-controlling steady-state solids were identified and the speciation and/or oxidation states present in the supernatant solutions were determined. The neptunium solubility decreased with increasing temperature and pH. Plutonium concentrations decreased with increasing temperature and showed no trend with pH. The americium solutions showed no clear solubility trend with increasing temperature and increasing pH

  16. Method of processing plutonium and uranium solution

    International Nuclear Information System (INIS)

    Solutions of plutonium nitrate solutions and uranyl nitrate recovered in the solvent extraction step in reprocessing plants and nuclear fuel production plants are applied with low temperature treatment by means of freeze-drying under vacuum into residues containing nitrates, which are denitrated under heating and calcined under reduction into powders. That is, since complicate processes of heating, concentration and dinitration conducted so far for the plutonium solution and uranyl solution are replaced with one step of freeze-drying under vacuum, the process can be simplified significantly. In addition, since the treatment is applied at low temperature, occurrence of corrosion for the material of evaporation, etc. can be prevented. Further, the number of operators can be saved by dividing the operations into recovery of solidification products, supply and sintering of the solutions and vacuum sublimation. Further, since nitrates processed at a low temperature are powderized by heating dinitration, the powderization step can be simplified. The specific surface area and the grain size distribution of the powder is made appropriate and it is possible to obtain oxide powders of physical property easily to be prepared into pellets. (N.H.)

  17. Extraction chromatographic recovery of americium from acidic raffinate solutions using CMPO adsorbed on Chromosorb-102

    International Nuclear Information System (INIS)

    Microgram amounts of americium have been separated and purified from large amounts of uranium present in effluent solutions resulting from the anion-exchange columns during the purification and recovery of plutonium by using TBP extraction followed by extraction chromatography using CMPO adsorbed on Chromosorb-102. (author). 4 refs., 1 tab

  18. Distribution of plutonium and americium in human and animal tissues after chronic exposures

    International Nuclear Information System (INIS)

    The distribution of plutonium in the tissues of a group of southern Finns was determined. Their Pu intake had been solely from fallout via inhalation. A group of northern Finns was also studied. They obtain most of the Pu from inhalation, but also some from their diet which is rich in reindeer liver. Reindeer obtain large amounts of transuranium elements in their natural winter diet, which mainly consists of lichen. Pu-239, 240 and Am-241 were also analyzed in elk because it is closely related to reindeer but does not feed on lichen. It was found that much of the Am-241 in reindeer tissues is due to ingrowth from Pu-241 in the animal. The aim of this study to establish whether this situation is also true for the human bone. (H.K.)

  19. Development of analytical method for the determination of uranium in presence of plutonium by complexometric titration

    International Nuclear Information System (INIS)

    Uranium determination in the plant samples by conventional Davies Gray method generates radioactive wastes bearing plutonium. Uranium determination by complexometric titration using 2,6 Pyridine di-carboxylic acid (PDCA) as titrant is a well established. It had been observed that the presence of plutonium increases bias in the determination of uranium. Hence for the Plutonium bearing process samples such as strip products, feed samples of the Plutonium reconversion, Uranium estimation through PDCA method is optimized by employing ethylene diamine tetra acetic (EDTA) as the masking agent. It is found that beyond 0.075 m moles of EDTA bias for determination of uranium increased; therefore EDTA of 0.05 mmoles is maintained in the titration medium. The precision for determination of uranium in absence of Pu for a range of 3-10 mg was found be less than 0.5 %. A bias of 6 % is observed for uranium determination with two different ratios of U and Pu. (author)

  20. Molten salt extraction (MSE) of americium from plutonium metal in CaCl2-KCl-PuCl3 and CaCl2-PuCl3 salt systems

    International Nuclear Information System (INIS)

    Molten salt extraction (MSE) of americium-241 from reactor-grade plutonium has been developed using plutonium trichloride salt in stationary furnaces. Batch runs with oxidized and oxide-free metal have been conducted at temperature ranges between 750 and 945C, and plutonium trichloride concentrations from one to one hundred mole percent. Salt-to-metal ratios of 0.10, 0.15, and 0 30 were examined. The solvent salt was either eutectic 74 mole percent CaCl2 endash 26 mole percent KCl or pure CaCl2. Evidence of trivalent product americium, and effects of temperature, salt-to-metal ratio, and oxide contamination on the americium extraction efficiency are given. 24 refs, 20 figs, 13 tabs

  1. Study Progress of On-line Monitoring Device for Uranium and Plutonium by XRF

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    An X-ray fluorescence spectrometer was designed and set up, which was used to determine uranium and plutonium on-line in reprocessing process stream. Uranium in aqueous and organic phase, plutonium in aqueous were measured by using the device,

  2. Recovery of plutonium and americium from laboratory acidic waste solutions using tri-n-octylamine and octylphenyl-N-N- diisobutylcarbamoylmethylphosphine oxide.

    Science.gov (United States)

    Michael, K M; Rizvi, G H; Mathur, J N; Kapoor, S C; Ramanujam, A; Iyer, R H

    1997-11-01

    Plutonium from acidic waste solutions has been recovered quantitatively using tri-n-octylamine (TnOA) in xylene and americium using a mixture of octylphenyl-N-N- diisobutylcarbamoylmethylphosphine oxide (CMPO) and TBP in dodecane by extraction and extraction chromatographic methods. The Pu ( IV ) TnOA species extracted into the organic phase from higher nitric acid concentrations has been confirmed as (R(3)NH)(2)Pu(NO(3))(6) (where R(3)N = TnOA by employing slope analysis as well as spectrophotometric studies. PMID:18966958

  3. Uptake of curium (244Cm) by five benthic marine species (Arenicola marina, Cerastoderma edule, Corophium volutator, Nereis diversicolor and Scrobicularia plana): comparison with americium and plutonium

    International Nuclear Information System (INIS)

    Curium (244Cm) uptake from contaminated sea water was studied in five benthic marine species: two bivalve molluscs (Scrobicularia plana and Cerastoderma edule), two polychaete annelids (Arenicola marina and Nereis diversicolor) and one amphidpod crustacean (Corophium volutator). The concentrations in the whole organisms relative to the concentration in the sea water (concentration factors) were: 700 for the amphipods (after 11 d of accumulation), 140 for the cockles (after 28 d), 80 for the scrobicularia (after 23d) and approx. 30 for the two annelids (after > 20 d). All species except S. plana accumulated americium and curium similarly; S. plana accumulated similar amounts of curium and plutonium. (author)

  4. Uptake of curium (/sup 244/Cm) by five benthic marine species (Arenicola marina, Cerastoderma edule, Corophium volutator, Nereis diversicolor and Scrobicularia plana): comparison with americium and plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Miramand, P.; Germain, P.; Arzur, J.C.

    1987-01-01

    Curium (/sup 244/Cm) uptake from contaminated sea water was studied in five benthic marine species: two bivalve molluscs (Scrobicularia plana and Cerastoderma edule), two polychaete annelids (Arenicola marina and Nereis diversicolor) and one amphidpod crustacean (Corophium volutator). The concentrations in the whole organisms relative to the concentration in the sea water (concentration factors) were: 700 for the amphipods (after 11 d of accumulation), 140 for the cockles (after 28 d), 80 for the scrobicularia (after 23d) and approx. 30 for the two annelids (after > 20 d). All species except S. plana accumulated americium and curium similarly; S. plana accumulated similar amounts of curium and plutonium.

  5. Photon attenuation properties of some thorium, uranium and plutonium compounds

    Energy Technology Data Exchange (ETDEWEB)

    Singh, V. P.; Badiger, N. M. [Karnatak University, Department of Physics, Dharwad-580003, Karnataka (India); Vega C, H. R., E-mail: kudphyvps@rediffmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2015-10-15

    Mass attenuation coefficients, effective atomic numbers, effective electron densities for nuclear materials; thorium, uranium and plutonium compounds have been studied. The photon attenuation properties for the compounds have been investigated for partial photon interaction processes by photoelectric effect, Compton scattering and pair production. The values of these parameters have been found to change with photon energy and interaction process. The variations of mass attenuation coefficients, effective atomic number and electron density with energy are shown graphically. Moreover, results have shown that these compounds are better shielding and suggesting smaller dimensions. The study would be useful for applications of these materials for gamma ray shielding requirement. (Author)

  6. Photon attenuation properties of some thorium, uranium and plutonium compounds

    International Nuclear Information System (INIS)

    Mass attenuation coefficients, effective atomic numbers, effective electron densities for nuclear materials; thorium, uranium and plutonium compounds have been studied. The photon attenuation properties for the compounds have been investigated for partial photon interaction processes by photoelectric effect, Compton scattering and pair production. The values of these parameters have been found to change with photon energy and interaction process. The variations of mass attenuation coefficients, effective atomic number and electron density with energy are shown graphically. Moreover, results have shown that these compounds are better shielding and suggesting smaller dimensions. The study would be useful for applications of these materials for gamma ray shielding requirement. (Author)

  7. Distribution of plutonium and americium beneath a 33-year-old liquid waste disposal site

    International Nuclear Information System (INIS)

    The distribution of Pu, 241Am, and water in Bandelier Tuff beneath a former liquid waste disposal site at Los Alamos was investigated. The waste use history of the site was described, as well as the previous field and laboratory studies of radionuclide migration performed at this site. One of the absorption beds studied had 20.5 m of water added to it in 1961 in an aggressive attempt to change the distribution of radionuclides in the tuff beneath the bed. Plutonium and 241Am were detected to sampling depths of 30 m in this bed, but only found to depths of 6.5 to 13.41 m in an adjacent absorption bed (bed 2) not receiving additional water in 1961. After 17 yr of migration of the slug water added to bed 1, 0.3 to 5.1% of the Pu inventory and 3.0 to 49.6% of the 241Am inventory was mobilized within the 30-m sampling depth, as less than one column volume of water moved through the tuff profile under the bed. The results of similar lab and field studies performed since 1953 were compared with our 1978 data and site hydrologic data was used as a time marker to estimate how fast radionuclide migration occurred in the tuff beneath absorption bed 1. Most of the radionuclide migration appeared to have occurred within 1 yr of the 20.5-m water leaching in 1961. 16 references, 2 figures, 4 tables

  8. Elemental bio-imaging of thorium, uranium, and plutonium in tissues from occupationally exposed former nuclear workers.

    Science.gov (United States)

    Hare, Dominic; Tolmachev, Sergei; James, Anthony; Bishop, David; Austin, Christine; Fryer, Fred; Doble, Philip

    2010-04-15

    Internal exposure from naturally occurring radionuclides (including the inhaled long-lived actinides (232)Th and (238)U) is a component of the ubiquitous background radiation dose (National Council on Radiation Protection and Measurements. Ionizing radiation exposure of the population of the United States; NCRP Report No. 160; NCRP: Bethesda, MD, 2009). It is of interest to compare the concentration distribution of these natural alpha-emitters in the lungs and respiratory lymph nodes with those resulting from occupational exposure, including exposure to anthropogenic plutonium and depleted and enriched uranium. This study examines the application of laser ablation-inductively coupled plasma-mass spectrometry (LA-ICPMS) to quantifying and visualizing the mass distribution of uranium and thorium isotopes from both occupational and natural background exposure in human respiratory tissues and, for the first time, extends this application to the direct imaging of plutonium isotopes. Sections of lymphatic and lung tissues taken from deceased former nuclear workers with a known history of occupational exposure to specific actinide elements (uranium, plutonium, or americium) were analyzed by LA-ICPMS. Using a previously developed LA-ICPMS protocol for elemental bio-imaging of trace elements in human tissue and a new software tool, we generated images of thorium ((232)Th), uranium ((235)U and (238)U), and plutonium ((239)Pu and (240)Pu) mass distributions in sections of tissue. We used a laboratory-produced matrix-matched standard to quantify the (232)Th, (235)U, and (238)U concentrations. The plutonium isotopes (239)Pu and (240)Pu were detected by LA-ICPMS in 65 mum diameter localized regions of both a paratracheal lymph node and a sample of lung tissue from a person who was occupationally exposed to refractory plutonium (plutonium dioxide). The average (overall) (239)Pu concentration in the lymph node was 39.2 ng/g, measured by high purity germanium (HPGe) gamma

  9. Kinetics of gaseous atoms in uranium plutonium mixed oxide

    International Nuclear Information System (INIS)

    The kinetics of fission gas atoms in nuclear fuel are much different from those discussed in academic fields such as statistical mechanics and/or statistical thermodynamics established by Ludwig Boltzmann. The field of movement should have unique microstructure such as pore, grain boundary, and especially plutonium enriched zone (Pu spot) in case of Uranium Plutonium Mixed Oxide (MOX). The mechanism of fission gas release has been studied by many researchers for these nearly fifty years since Booth first set down the simple approximations to fission gas release from irradiated UO2. However, there is still no general agreement what mechanisms are responsible for observed coalescence of intergranular bubbles during their growth. It might be due to the complexity of the kinetics of gaseous atoms in nuclear fuel pellet having unique microstructure, temperature gradient and irradiation damage. The objectives of this paper are to introduce brief review of fundamental process of kinetics of gaseous atom in nuclear fuel and to develop effective fission gas release model. The fission gas release mechanism for MOX fuel which has heterogeneous microstructure including Pu spot is introduced based on the result of post-irradiation examination with MOX having high plutonium content. (authors)

  10. New Synthetic Methods and Structure-Property Relationships in Neptunium, Plutonium, and Americium Borates. Final report

    International Nuclear Information System (INIS)

    The past three years of support by the Heavy Elements Chemistry Program have been highly productive in terms of advanced degrees awarded, currently supported graduate students, peer-reviewed publications, and presentations made at universities, national laboratories, and at international conferences. Ph.D. degrees were granted to Shuao Wang and Juan Diwu, who both went on to post-doctoral appointments at the Glenn T. Seaborg Center at Lawrence Berkeley National Laboratory with Jeff Long and Ken Raymond, respectively. Pius Adelani completed his Ph.D. with me and is now a post-doc with Peter C. Burns. Andrea Alsobrook finished her Ph.D. and is now a post-doc at Savannah River with Dave Hobbs. Anna Nelson completed her Ph.D. and is now a post-doc with Rod Ewing at the University of Michigan. As can be gleaned from this list, students supported by the Heavy Elements Chemistry grant have remained interested in actinide science after leaving my program. This follows in line with previous graduates in this program such as Richard E. Sykora, who did his post-doctoral work at Oak Ridge National Laboratory with R. G. Haire, and Amanda C. Bean, who is a staff scientist at Los Alamos National Laboratory, and Philip M. Almond and Thomas C. Shehee, who are both staff scientists at Savannah River National Laboratory, Gengbang Jin who is a staff scientist at Argonne National Lab, and Travis Bray who has been a post-doc at both LBNL and ANL. Clearly this program is serving as a pipe-line for students to enter into careers in the national laboratories. About half of my students depart the DOE complex for academia or industry. My undergraduate researchers also remain active in actinide chemistry after leaving my group. Dan Wells was a productive undergraduate of mine, and went on to pursue a Ph.D. on uranium and neptunium chalcogenides with Jim Ibers at Northwestern. After earning his Ph.D., he went directly into the nuclear industry

  11. New Synthetic Methods and Structure-Property Relationships in Neptunium, Plutonium, and Americium Borates. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Albrecht-Schmitt, Thomas Edward

    2013-09-14

    The past three years of support by the Heavy Elements Chemistry Program have been highly productive in terms of advanced degrees awarded, currently supported graduate students, peer-reviewed publications, and presentations made at universities, national laboratories, and at international conferences. Ph.D. degrees were granted to Shuao Wang and Juan Diwu, who both went on to post-doctoral appointments at the Glenn T. Seaborg Center at Lawrence Berkeley National Laboratory with Jeff Long and Ken Raymond, respectively. Pius Adelani completed his Ph.D. with me and is now a post-doc with Peter C. Burns. Andrea Alsobrook finished her Ph.D. and is now a post-doc at Savannah River with Dave Hobbs. Anna Nelson completed her Ph.D. and is now a post-doc with Rod Ewing at the University of Michigan. As can be gleaned from this list, students supported by the Heavy Elements Chemistry grant have remained interested in actinide science after leaving my program. This follows in line with previous graduates in this program such as Richard E. Sykora, who did his post-doctoral work at Oak Ridge National Laboratory with R. G. Haire, and Amanda C. Bean, who is a staff scientist at Los Alamos National Laboratory, and Philip M. Almond and Thomas C. Shehee, who are both staff scientists at Savannah River National Laboratory, Gengbang Jin who is a staff scientist at Argonne National Lab, and Travis Bray who has been a post-doc at both LBNL and ANL. Clearly this program is serving as a pipe-line for students to enter into careers in the national laboratories. About half of my students depart the DOE complex for academia or industry. My undergraduate researchers also remain active in actinide chemistry after leaving my group. Dan Wells was a productive undergraduate of mine, and went on to pursue a Ph.D. on uranium and neptunium chalcogenides with Jim Ibers at Northwestern. After earning his Ph.D., he went directly into the nuclear industry.

  12. The Biokinetic Model of Americium

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    To improve in vivo measurements for detecting internal exposure from transuranium radio nuclides, such as neptunium, plutonium, americium, the bioknetic model was studied. According to ICRP report (1993, 1995, 1997) and other research, the

  13. The spectrographic analysis of plutonium oxide or mixed plutonium oxide/uranium oxide fuel pellets by the dried residue technique

    International Nuclear Information System (INIS)

    An emission spectrographic method for the quantitative determination of metallic impurities in plutonium oxide and mixed plutonium oxide/uranium oxide is described. The fuel is dissolved in nitric acid and the plutonium and/or uranium extracted with tributyl phosphate. A small aliquot of the aqueous residue is dried on a 'mini' pyrolitic graphite plate and excited by high voltage AC spark in an oxygen atmosphere. Spectra are recorded in a region which has been specially selected to record simultaneously lines of boron and cadmium in the 2nd order and all the other elements of interest in the 1st order. Indium is used as an internal standard. The excitation of very small quantities of the uraniumm/plutonium free residue by high voltage spark, together with three separate levels of containment reduce the hazards to personnel and the environment to a minimum with limited effect on sensitivity and accuracy of the results. (auth)

  14. Performance evaluation of indigenous controlled potential coulometer for the determination of uranium and plutonium

    International Nuclear Information System (INIS)

    We have carried out performance evaluation of indigenously manufactured controlled potential coulometer for the determination of uranium and plutonium respectively in Rb2U(SO4)3 and K4Pu(SO4)4 chemical assay standards. The coulometric results obtained on uranium determination showed an insignificant difference as compared with the biamperometric results at 95% and 99.9% confidence levels while for plutonium determination showed a difference of -0.4% at 95% with respect to expected value. The results obtained show that indigenous coulometer is suitable for uranium and plutonium determination in chemical assay standards. (author)

  15. Recycling of plutonium and uranium in water reactor fuels

    International Nuclear Information System (INIS)

    The purpose of the meeting was to make a review of the present knowledge relevant to plutonium and uranium recycling, MOX fuel, on-going programmes, today's industrial capabilities and future plans for development. For countries with commitments to reprocessing, MOX fuel is attractive and will be more so as discharge burnups increase and as the time between discharge and reprocessing optimized. Fabrication experience on MOX fuel has accumulated for many years in several countries and one has been able to cope with the extension of capacities of the plants, as required by MOX fuel implementation, and with the requirements specific to massive use in power reactors. Standards fabrication processes have proven to be adaptable in large quantities and have yielded products satisfying all present specifications. A large body of irradiation experience for some time on various MOX and RepU materials. On the basis of a comparison with UO2, no adverse effect has been observed. Problems like isotopic homogeneity, solubility, alternative processes like gelation deserve further attention. It is encouraging to note that parameters linked to materials obtained by different fabrication routes can be taken into account by existing codes, to an extent similar to various UO2 fuels, provided an adequate data base is available. The fabrication capacities are the limiting factor for MOX penetration in reactors, where a 30 to 50% recycling rate is therefore sufficient. The use of plutonium in 100% MOX reactors or in more advanced reactors deserves more study. The increase of plutonium inventory may influence safety and licensing analysis, but all the safety criteria can be met. On the whole, the experience reported in this meeting pointed to a general consensus of the attractiveness of recycling and the already demonstrated ability of several countries to cope with all questions raised by MOX substitution of UO2 fuel. Refs, figs and tabs

  16. Uranium-plutonium carbide as an LMFBR advanced fuel

    International Nuclear Information System (INIS)

    Uranium-plutonium carbide offers an improved fuel system for advanced breeder reactors. The high thermal conductivity and density of carbide fuels permit superior breeding performance and high specific power operation. These advantages combine to increase plutonium production, reduce fuel cycle and power costs, and lower plant capital costs. The carbide advantages are obtained at conservative fuel sytem design and operating conditions. Carbide fabrication technology has been demonstrated by the production of quality-assured fuel elements for irradiation testing. The carbide irradiation test program has demonstrated that high burnup can be achieved with several designs and that the consequences of postulated off-normal operating events are benign. Design bases to support helium- and sodium-bonded carbide fuel pin test irradiations in the Fast Flux Test Facility have been developed in the Experimental Breeder Reactor-II and the Transient Reactor irradiation experiments. Important issues regarding safety, reprocessing, and commercial-scale fabrication remain to be addressed in the continuing development of carbide fuels. Fiscal and historical circumstances have combined to preclude this development. This report reviews these circumstances and the state of the technology in general and advances a rationale for why development should be continued

  17. Main radiation characteristics of spent nuclear uranium and uranium-plutonium fuel at long-term storage

    International Nuclear Information System (INIS)

    Comparison of radiotoxicity and decay heat power of spent uranium and uranium - plutonium nuclear fuel at long-term storage is performed. The contributions of the most important nuclides are determined. Their prime extraction and transmutation permits to reduce decay heat power and radiotoxicity of wastes staying in storage.(author)

  18. Separation Techniques for Uranium and Plutonium at Trace Levels for the Thermal Ionization Mass Spectrometric Determination

    International Nuclear Information System (INIS)

    This report describes the state of the art and the progress of the chemical separation and purification techniques required for the thermal ionization mass spectrometric determination of uranium and plutonium in environmental samples at trace or ultratrace levels. Various techniques, such as precipitation, solvent extraction, extraction chromatography, and ion exchange chromatography, for separation of uranium and plutonium were evaluated. Sample preparation methods and dissolution techniques for environmental samples were also discussed. Especially, both extraction chromatographic and anion exchange chromatographic procedures for uranium and plutonium in environmental samples, such as soil, sediment, plant, seawater, urine, and bone ash were reviewed in detail in order to propose some suitable methods for the separation and purification of uranium and plutonium from the safeguards environmental or swipe samples. A survey of the IAEA strengthened safeguards system, the clean room facility of IAEA's NWAL(Network of Analytical Laboratories), and the analytical techniques for safeguards environmental samples was also discussed here

  19. Dioctyl butyramide and dioctyl isobutyramide as extractants for uranium(VI) and plutonium(IV)

    International Nuclear Information System (INIS)

    Two isomeric monoamides, dioctyl butyramide (DOBA) and dioctyl is isobutyramide (DOIBA) were synthesized for extracting uranium(VI) and plutonium(IV) from aqueous nitric acid medium into various diluents such as n-dodecane, tertiary butyl benzene and xylene. DOBA extracted uranium(VI) and plutonium(IV) efficiently whereas DOIBA extracted uranium(VI) with negligible extraction for plutonium(IV). Both these cations were extracted as their disolvates. The thermodynamic parameters involved in the extraction determined by the temperature variation method indicated the reactions in all cases to be enthalpy favoured and entropy disfavoured. Possibility of separating micrograms of plutonium(IV) from macro quantities of uranium(VI) using the mixture of these amides was explored. (author). 15 refs., 6 figs., 1 tab

  20. Separation Techniques for Uranium and Plutonium at Trace Levels for the Thermal Ionization Mass Spectrometric Determination

    Energy Technology Data Exchange (ETDEWEB)

    Suh, M. Y.; Han, S. H.; Kim, J. G.; Park, Y. J.; Kim, W. H

    2005-12-15

    This report describes the state of the art and the progress of the chemical separation and purification techniques required for the thermal ionization mass spectrometric determination of uranium and plutonium in environmental samples at trace or ultratrace levels. Various techniques, such as precipitation, solvent extraction, extraction chromatography, and ion exchange chromatography, for separation of uranium and plutonium were evaluated. Sample preparation methods and dissolution techniques for environmental samples were also discussed. Especially, both extraction chromatographic and anion exchange chromatographic procedures for uranium and plutonium in environmental samples, such as soil, sediment, plant, seawater, urine, and bone ash were reviewed in detail in order to propose some suitable methods for the separation and purification of uranium and plutonium from the safeguards environmental or swipe samples. A survey of the IAEA strengthened safeguards system, the clean room facility of IAEA's NWAL(Network of Analytical Laboratories), and the analytical techniques for safeguards environmental samples was also discussed here.

  1. Production of americium isotopes in France

    International Nuclear Information System (INIS)

    The program of productions of americium 241 and 243 isotopes is based respectively on the retreatment of aged plutonium alloys or plutonium dioxide and on the treatment of plutonium targets irradiated either in CELESTIN reactors for Pu-Al alloys or OSIRIS reactor for plutonium 242 dioxide. All the operations, including americium final purifications, are carried out in hot cells equipped with remote manipulators. The chemical processes are based on the use of extraction chromatography with hydrophobic SiO2 impregnated with extracting agents. Plutonium targets and aged plutonium alloys are dissolved in nitric acid using conventional techniques while plutonium dioxide dissolutions are performed routine at 300 grams scale with electrogenerated silver II in 4M HNO3 at room temperature. The separation between plutonium and americium is performed by extraction of Pu(IV) either on TBP/SiO2 or TOAHNO3/SiO2 column. Americium recovery from waste streams rid of plutonium is realized by chromatographic extraction of Am(III) using mainly TBP and episodically DHDECMP as extractant. The final purification of both americium isotopes uses the selective extraction of Am(VI) on HDDiBMP/SiO2 column at 60 grams scale. Using the overall process a total amount of 1000 grams of americium 241 and 100 grams of americium 243 has been produced nowadays and the AmO2 final product indicates a purity better than 98.5%

  2. Technique for radiation monitoring of uranium and plutonium in water solutions

    International Nuclear Information System (INIS)

    Nuclear materials like uranium and plutonium are widely used in the form of different water solutions, i.e. processing media, cores of solution reactors, liquid radioactive wastes from industrial facilities, etc. Furthermore, isotopes of these elements pose serious danger from the standpoint of natural environments including natural waters. Titanium hydroxide is a material found to be one of the best for uranium extraction from water solutions. But publications give few data on the titanium hydroxide application in uranium detection procedures and have no data on sorption characteristics of titanium hydroxide relative to plutonium. So, systematic investigations into sorption characteristics of titanium hydroxide with respect to uranium and plutonium, as well as investigations how the sorption process depends on various factors were performed to study feasibility of titanium hydroxide application for efficient extraction of uranium and plutonium isotopes from aqueous solutions. Compositional sorbent (grade TG-TTs), i.e. thin-layered titanium hydroxide (10-15 μg/cm2 titanium content) chemically applied onto the cellulose triacetate film was studied as a sorbent to extract uranium and plutonium from aqueous media. Investigations into uranium and plutonium sorption and how it depends on the solution acidity demonstrated that uranium extraction from aqueous media is most complete when solution acidity is pH 5-8 and plutonium extraction - when solution acidity is pH 3-6. Investigation how uranium sorption depends on time demonstrated that at room temperature the sorption process is time-consuming. So, increase of the solution temperature is one of the main factors that increases sorption rate. Sorption at 70-900? ensures that the main part of uranium (about 70%) is adsorbed during 45 minutes. The obtained data on the sorbent capacity as for uranium and plutonium show that their qualitative extraction is ensured if the 'solution volume-to-sorbent area' ratio is 1 ml:1cm2

  3. Measurement of plutonium and uranium isotopic abundances by gamma-ray spectrometry

    International Nuclear Information System (INIS)

    The isotopic composition of plutonium and uranium is needed for purposes of sample confirmation, or for interpreting results from calorimeters or neutron-coincidence measurement instruments to determine nuclear material mass. The authors have developed measurement methods and computer codes utilizing high-resolution gamma-ray spectrometry to measure the relative isotopic abundances of plutonium and uranium in various forms nondestructively. The computer codes, known as MGA and MGAU, have unique analysis methodologies that the authors briefly describe in this paper

  4. Measurement of plutonium and uranium isotopic abundances by gamma-ray spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Ruhter, W.D. [Lawrence Livermore National Lab., CA (United States); Gunnink, R. [Gunnink (Ray), Fremont, CA (United States)

    1996-02-01

    The isotopic composition of plutonium and uranium is needed for purposes of sample confirmation, or for interpreting results from calorimeters or neutron-coincidence measurement instruments to determine nuclear material mass. The authors have developed measurement methods and computer codes utilizing high-resolution gamma-ray spectrometry to measure the relative isotopic abundances of plutonium and uranium in various forms nondestructively. The computer codes, known as MGA and MGAU, have unique analysis methodologies that the authors briefly describe in this paper.

  5. On line spectrophotometry with optical fibers. Application to uranium-plutonium separation in a spent fuel reprocessing plant

    International Nuclear Information System (INIS)

    Optimization of mixer-settler operation for uranium-plutonium separation in the Purex process can be obtained by remote spectrophotometry with optical fibers. Data acquisition on uranium VI, uranium IV and plutonium III is examined in function of acidity and nitrate content of the solution. Principles for on line multicomponent monitoring and mathematical modelization of the measurements are described

  6. Study of the extraction and the purification of americium and trivalent actinides contained in effluents with supported liquid membranes

    International Nuclear Information System (INIS)

    The supported liquid membrane technique is studied and developed for americium recovery from uranium or plutonium matrices and decontamination of liquid radioactive wastes. First tests on uranium-nickel solutions with a flat membrane showed the easiness of the operation and the efficiency of the process. Acid-resistant (10 N), interchangeable elements with hollow fibers, are developed and also a computerized automatic device. The different tests on americium solutions demonstrate the feasibility and the reliability of the system. Influence of various parameters on transfer kinetics is investigated

  7. SORPTION OF URANIUM, PLUTONIUM AND NEPTUNIUM ONTO SOLIDS PRESENT IN HIGH CAUSTIC NUCLEAR WASTE STORAGE TANKS

    Energy Technology Data Exchange (ETDEWEB)

    Oji, L; Bill Wilmarth, B; David Hobbs, D

    2008-05-30

    Solids such as granular activated carbon, hematite and sodium phosphates, if present as sludge components in nuclear waste storage tanks, have been found to be capable of precipitating/sorbing actinides like plutonium, neptunium and uranium from nuclear waste storage tank supernatant liqueur. Thus, the potential may exists for the accumulation of fissile materials in such nuclear waste storage tanks during lengthy nuclear waste storage and processing. To evaluate the nuclear criticality safety in a typical nuclear waste storage tank, a study was initiated to measure the affinity of granular activated carbon, hematite and anhydrous sodium phosphate to sorb plutonium, neptunium and uranium from alkaline salt solutions. Tests with simulated and actual nuclear waste solutions established the affinity of the solids for plutonium, neptunium and uranium upon contact of the solutions with each of the solids. The removal of plutonium and neptunium from the synthetic salt solution by nuclear waste storage tank solids may be due largely to the presence of the granular activated carbon and transition metal oxides in these storage tank solids or sludge. Granular activated carbon and hematite also showed measurable affinity for both plutonium and neptunium. Sodium phosphate, used here as a reference sorbent for uranium, as expected, exhibited high affinity for uranium and neptunium, but did not show any measurable affinity for plutonium.

  8. The development of uranium-plutonium glass standards for hybrid K-edge densitometry

    International Nuclear Information System (INIS)

    Hybrid K-Edge Densitometry (HKED) is a valuable technique for the rapid determination of uranium and plutonium concentration in spent fuel input solution. The IAEA plan to install HKED equipment at Rokkasho and to use it as the primary means of measurement of uranium and plutonium in input solution and process tanks. Standards are required for use in the quality control of measurements. For on-site application, standards in the form of a stable matrix offer distinct advantages over solution standards, which can deteriorate due to evaporation and radiolysis in addition to requiring careful handling. Within the framework of Task UK A01057 of the UK Support Programme to the IAEA, work has been undertaken to prepare actinide glass standards. Glass compositions have been reviewed and glasses selected that should dissolve the required amounts of uranium and plutonium: up to 300g/litre uranium and 3g/litre plutonium. Six glasses with different concentrations of uranium and plutonium have been prepared. Some of the glasses have been cut and polished to fit within 1cm optical cells, the specified geometry for the quality control standards. Attempts to characterise the glasses using isotope-dilution mass spectrometry have been unsuccessful. Incomplete dissolution is believed to be the cause of the low and variable recoveries of both uranium and plutonium compared with make-up values. Some of the standards have been shipped to the Safeguards Analytical Laboratory for trial measurement by HKED. However, the specification for the geometry of the standard has subsequently changed, making the current actinide glasses obsolete. The IAEA now require HKED standards to fit within sample jugs designed for use in the Rokkasho sample pneumatic transfer system. It is unlikely that these standards can be prepared from actinide glasses, and it is recommended that the IAEA investigate the possibility of using set resins containing uranium and plutonium instead of glass. (author)

  9. Preliminary results from uranium/americium affinity studies under experimental conditions for cesium removal from NPP ''Kozloduy'' simulated wastes solutions

    International Nuclear Information System (INIS)

    We use the approach described by Westinghouse Savannah River Company using ammonium molybdophosphate (AMP) to remove elevated concentrations of radioactive cesium to facilitate handling waste samples from NPP Kozloduy. Preliminary series of tests were carried out to determine the exact conditions for sufficient cesium removal from five simulated waste solutions with concentrations of compounds, whose complexing power complicates any subsequent processing. Simulated wastes solutions contain high concentrations of nitrates, borates, H2C2O4, ethylenediaminetetraacetate (EDTA) and Citric acid, according to the composition of the real waste from the NPP. On this basis a laboratory treatment protocol was created. This experiment is a preparation for the analysis of real waste samples. In this sense the results are preliminary. Unwanted removal of non-cesium radioactive species from simulated waste solutions was studied with gamma spectrometry with the aim to find a compromise between on the one hand the AMP effectiveness and on the other hand unwanted affinity to AMP of Uranium and Americium. Success for the treatment protocol is defined by proving minimal uptake of U and Am, while at the same time demonstrating good removal effectiveness through the use of AMP. Uptake of U and Am were determined as influenced by oxidizing agents at nitric acid concentrations, proposed by Savannah River National laboratory. It was found that AMP does not significantly remove U and Am when concentration of oxidizing agents is more than 0.1M for simulated waste solutions and for contact times inherent in laboratory treatment protocol. Uranium and Americium affinity under experimental conditions for cesium removal were evaluated from gamma spectrometric data. Results are given for the model experiment and an approach for the real waste analysis is chosen. Under our experimental conditions simulated wastes solutions showed minimal affinity to AMP when U and Am are most probably in the

  10. Reaction of uranium and plutonium carbides with austenitic steels

    International Nuclear Information System (INIS)

    The reaction of uranium and plutonium carbides with austenitic steels has been studied between 650 and 1050 deg. C using UC, steel and (UPu)C, steel diffusion couples. The steels are of the type CN 18.10 with or without addition of molybdenum. The carbides used are hyper-stoichiometric. Tests were also carried out with UCTi, UCMo, UPuCTi and UPuCMo. Up to 800 deg. C no marked diffusion of carbon into stainless steel is observed. Between 800 and 900 deg. C the carbon produced by the decomposition of the higher carbides diffuses into the steel. Above 900 deg. C, decomposition of the monocarbide occurs according to a reaction which can be written schematically as: (U,PuC) + (Fe,Ni,Cr) → (U,Pu) Fe2 + Cr23C6. Above 950 deg. C the behaviour of UPuCMo and that of the titanium (CN 18.12) and nickel (NC 38. 18) steels is observed to be very satisfactory. (author)

  11. Nuclear data needs for uranium-plutonium fuel cycle development

    International Nuclear Information System (INIS)

    Potential needs of nuclear data have been surveyed through the uranium-plutonium fuel cycle. A wide variety of data were used in operation of fuel cycle facilities and its relating activities. Those were decay data, cross section data and fission-product yield data. Although the nuclear data were used mostly as physical constants in analyzing results of measurements and calculations made, the users did not give a first priority to their requirements on the nuclear data. It is necessary to disseminate uniquely evaluated data for securing uniformity of data-treatment basis of the users. A covariance file of the evaluated cross section data was requested by workers involved in reactor dosimetry studies and in actinides incineration studies for waste management. Also, in the various parts of the fuel cycle equally required were some evaluated computer codes including reasonably organized nuclear data set which enable one to predict the amounts of actinides and fission-product nuclides in irradiated nuclear fuels. (Auth.)

  12. Gravimetric determination of carbon in uranium-plutonium carbide materials

    International Nuclear Information System (INIS)

    A gravimetric method for determining carbon in uranium-plutonium carbide materials was developed to analyze six samples simultaneously. The samples are burned slowly in an oxygen atmosphere at approximately 9000C, and the gases generated are passed through Schuetze's oxidizing reagent (iodine pentoxide on silica gel) to assure quantitative oxidation of the CO to CO2. The CO2 is collected on Ascarite and weighed. This method was tested using a tungsten carbide reference material (NBS-SRM-276) and a (U,Pu)C sample. For 42 analyses of the tungsten carbide, which has a certified carbon content of 6.09%, an average value of 6.09% was obtained with a standard deviation of 0.017% or a relative standard deviation of 0.28%. For 17 analyses of the (U,Pu)C sample, an average carbon content of 4.97% was found with a standard deviation of 0.012% or a relative standard deviation of 0.24%

  13. Study of reactions between uranium-plutonium mixed oxide and uranium nitride and between uranium oxide and uranium nitride; Etude des reactions entre l`oxyde mixte d`uranium-plutonium et le nitrure d`uranium et entre l`oxyde d`uranium et le nitrure d`uranium

    Energy Technology Data Exchange (ETDEWEB)

    Lecraz, C.

    1993-06-11

    A new type of combustible elements which is a mixture of uranium nitride and uranium-plutonium oxide could be used for Quick Neutrons Reactors. Three different studies have been made on the one hand on the reactions between uranium nitride (UN) and uranium-plutonium mixed oxide (U,Pu)O{sub 2}, on the other hand on these between UN and uranium oxide UO{sub 2}. They show a sizeable reaction between nitride and oxide for the studied temperatures range (1573 K to 1973 K). This reaction forms a oxynitride compound, MO{sub x} N{sub y} with M=U or M=(U,Pu), whose crystalline structure is similar to oxide`s. Solubility of nitride in both oxides is studied, as the reaction kinetics. (TEC). 32 refs., 48 figs., 22 tabs.

  14. Survey of plutonium and uranium atom ratios and activity levels in Mortandad Canyon

    Energy Technology Data Exchange (ETDEWEB)

    Gallaher, B.M.; Benjamin, T.M.; Rokop, D.J.; Stoker, A.K.

    1997-09-22

    For more than three decades Mortandad Canyon has been the primary release area of treated liquid radioactive waste from the Los Alamos National Laboratory (Laboratory). In this survey, six water samples and seven stream sediment samples collected in Mortandad Canyon were analyzed by thermal ionization mass spectrometry (TIMS) to determine the plutonium and uranium activity levels and atom ratios. Be measuring the {sup 240}Pu/{sup 239}Pu atom ratios, the Laboratory plutonium component was evaluated relative to that from global fallout. Measurements of the relative abundance of {sup 235}U and {sup 236}U were also used to identify non-natural components. The survey results indicate the Laboratory plutonium and uranium concentrations in waters and sediments decrease relatively rapidly with distance downstream from the major industrial sources. Plutonium concentrations in shallow alluvial groundwater decrease by approximately 1000 fold along a 3000 ft distance. At the Laboratory downstream boundary, total plutonium and uranium concentrations were generally within regional background ranges previously reported. Laboratory derived plutonium is readily distinguished from global fallout in on-site waters and sediments. The isotopic ratio data indicates off-site migration of trace levels of Laboratory plutonium in stream sediments to distances approximately two miles downstream of the Laboratory boundary.

  15. Ingot formation using uranium dendrites recovered by electrolysis in LiCl-KCl-PuCl3-UCl3 melt

    International Nuclear Information System (INIS)

    Products on solid cathodes recovered by the metal pyrochemical processing were processed to obtain uranium ingot. Studies on process conditions of uranium formation, assay recovered uranium products and by-products and evaluation of mass balance were carried out. In these tests, it is confirmed that uranium ingots can be obtained with heating the products more than melting temperature of metal uranium under atmospheric pressure because adhered salt cover the uranium not to oxidize it during uranium cohering. Covered salt can be removed after ingot formation. Inside the ingot, there were a lump of uranium and dark brown colored dross was observed. Material balance of uranium is 77 ∼ 96%, that of plutonium is 71 ∼ 109%, and that of americium that is a volatile substance more than uranium and plutonium become 79 ∼ 119%. Volatilization of americium is very small under the condition of high temperature. (authors)

  16. Combined procedure using radiochemical separation of plutonium, americium and uranium radionuclides for alpha-spectrometry

    International Nuclear Information System (INIS)

    Radiochemical separation of Pu, Am and U was tested from synthetic solutions and evaporator concentrate samples from nuclear power plants for isolation of each of them for alpha-spectrometry analysis. The separation was performed by anion-exchange chromatography, extraction chromatography, using TRU resin, and precipitation techniques. The aim of the study was to develop a sensitive analytical procedure for the sequential determination of 242Pu, 238Pu, 239+240Pu, 241Am and 235,238U in radioactive wastes. 238Pu, 242Pu, 243Am and 232U were used as tracers. The measurements of α emitting radionuclides were performed by semiconductor detector that is used especially when spectrometric information is needed. For synthetic solutions the chemical recovery was based on associated iron concentration and was about 93%. (author)

  17. Kinetics of the oxidation-reduction reactions of uranium, neptunium, plutonium, and americium in aqueous solutions

    International Nuclear Information System (INIS)

    This is a review with about 250 references. Data for 240 reactions are cataloged and quantitative activation parameters are tabulated for 79 of these. Some empirical correlations are given. Twelve typical reactions are discussed in detail, along with the effects of self-irradiation and ionic strength. (U.S.)

  18. Analysis of americium-beryllium neutron source composition using the FRAM code

    Energy Technology Data Exchange (ETDEWEB)

    Hypes, P. A. (Philip A.); Bracken, D. S. (David S.); Sampson, Thomas E.; Taylor, W. A. (Wayne A.)

    2002-01-01

    The FRAM code was originally developed to analyze high-resolution gamma spectra from plutonium items. Its capabilities have since been expanded to include analysis of uranium spectra. The flexibility of the software also enables a capable spectroscopist to use FRAM to analyze spectra in which neither plutonium nor uranium is present in significant amounts. This paper documents the use of FRAM to determine the {sup 239}Pu/{sup 241}Am, {sup 243}Am/{sup 241}Am, {sup 237}Np/{sup 241}Am, and {sup 239}Np/{sup 241}Am ratios in americium-beryllium neutron sources. The effective specific power of each neutron source was calculated from the ratios determined by FRAM in order to determine the americium mass of each of these neutron sources using calorimetric assay. We will also discuss the use of FRAM for the general case of isotopic analysis of nonplutonium, nonuranium items.

  19. Uranium in the Nuclear Fuel Cycle: Creation of Plutonium (Invited)

    Science.gov (United States)

    Ewing, R. C.

    2009-12-01

    One of the important properties of uranium is that it can be used to “breed” higher actinides, particularly plutonium. During the past sixty years, more than 1,800 metric tonnes of Pu, and substantial quantities of the “minor” actinides, such as Np, Am and Cm, have been generated in nuclear reactors - a permanent record of nuclear power. Some of these transuranium elements can be a source of energy in fission reactions (e.g., 239Pu), a source of fissile material for nuclear weapons (e.g., 239Pu and 237Np), and of environmental concern because of their long-half lives and radiotoxicity (e.g., 239Pu and 237Np). In fact, the new strategies of the Advance Fuel Cycle Initiative (AFCI) are, in part, motivated by an effort to mitigate some of the challenges of the disposal of these long-lived actinides. There are two basic strategies for the disposition of these heavy elements: 1.) to “burn” or transmute the actinides using nuclear reactors or accelerators; 2.) to “sequester” the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. There has been substantial interest in the use of actinide-bearing minerals, such as zircon or isometric pyrochlore, A2B2O7 (A= rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage1. The radiation stability of these compositions is closely related to the structural distortions that can be accommodated for specific pyrochlore compositions and the electronic structure of the B-site cation. Recent developments in the understanding of the properties of heavy element solids have opened up new possibilities for the design of advanced nuclear fuels and waste forms.

  20. Water Solubility of Plutonium and Uranium Compounds and Residues at TA-55

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, Sean Douglas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Smith, Paul Herrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Jarvinen, Gordon D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Prochnow, David Adrian [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Schulte, Louis D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; DeBurgomaster, Paul Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Fife, Keith William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Rubin, Jim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Worl, Laura Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States

    2016-06-13

    Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that the following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U3O8 (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significant quantities of PuCl3, and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a commercially

  1. Water Solubility of Plutonium and Uranium Compounds and Residues at TA-55

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, Sean Douglas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Smith, Paul Herrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Jarvinen, Gordon D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Prochnow, David Adrian [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Schulte, Louis D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; DeBurgomaster, Paul Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Fife, Keith William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Rubin, Jim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Worl, Laura Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States

    2016-06-13

    Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that the following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U3O8 (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significant quantities of PuCl3, and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a

  2. Characteristics of plutonium burning MOX and uranium-free PWRs

    International Nuclear Information System (INIS)

    The large quantities of Pu currently accumulated worldwide could be reduced by employing 100% MOX LWR cores. A more effective way would be to use uranium-free fuel so that no new Pu is produced. The present paper compares the potential and the possible difficulties of plutonium burning PWR cores on the basis of two PWRs of different power. The first type consists of 100% MOX loadings. The second one employs a uranium-free fuel consisting of Pu in a neutronically inert matrix of ZrO2, along with a burnable poison (Er2O3) added to reduce the reactivity swing during burnup. The characteristics of the cores employing MOX or the U-free fuel have been studied via 3D-calculations for the equilibrium cycle and compared to real life reactor cycles which have been used as reference. For all cases appears possible to design a 4-region core with a natural cycle length of about 300 days. In the 100% MOX cores the Pu mass is reduced by about 30% of the initial Pu inventory. In the U-free core the reduction is twice as much, i.e. about 60%, because no new Pu is produced during burnup. The reactivity balance for HFP to HZP shows, on the one hand, that the 100% MOX cores need more effective control rods, and on the other hand, that the U-free core has larger reactivity shutdown margins than the reference one. The steamline break accident could be problematic in the MOX cores due to a large negative moderator temperature coefficient and a reduced value of the control rod due to the hard neutron spectrum. On the other hand, the Pu-Er core is expected to behave in about the same way as the reference UO2 core in this context. The rod ejection transient should not present any problem because the maximum inserted worth of a control rod is less than 0.4$. (author) 4 figs., 9 tabs., 11 refs

  3. Minutes of the 28th Annual Plutonium Sample Exchange Meeting. Part II: metal sample exchange

    International Nuclear Information System (INIS)

    Contents of this publication include the following list of participating laboratories; agenda; attendees; minutes of October 25 and 26 meeting; and handout materials supplied by speakers. The handout materials cover the following: statistics and reporting; plutonium - chemical assay 100% minus impurities; americium neptunium, uranium, carbon and iron data; emission spectroscopy data; plutonium metal sample exchange; the calorimetry sample exchange; chlorine determination in plutonium metal using phyrohydrolysis; spectrophotometric determination of 238-plutonium in oxide; plutonium measurement capabilities at the Savannah River Plant; and robotics in radiochemical laboratory

  4. Minutes of the 28th Annual Plutonium Sample Exchange Meeting. Part II: metal sample exchange

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    Contents of this publication include the following list of participating laboratories; agenda; attendees; minutes of October 25 and 26 meeting; and handout materials supplied by speakers. The handout materials cover the following: statistics and reporting; plutonium - chemical assay 100% minus impurities; americium neptunium, uranium, carbon and iron data; emission spectroscopy data; plutonium metal sample exchange; the calorimetry sample exchange; chlorine determination in plutonium metal using phyrohydrolysis; spectrophotometric determination of 238-plutonium in oxide; plutonium measurement capabilities at the Savannah River Plant; and robotics in radiochemical laboratory.

  5. Comparative XRPD and XAS study of the impact of the synthesis process on the electronic and structural environments of uranium-americium mixed oxides

    Science.gov (United States)

    Prieur, D.; Lebreton, F.; Martin, P. M.; Caisso, M.; Butzbach, R.; Somers, J.; Delahaye, T.

    2015-10-01

    Uranium-americium mixed oxides are potential compounds to reduce americium inventory in nuclear waste via a partitioning and transmutation strategy. A thorough assessment of the oxygen-to-metal ratio is paramount in such materials as it determines the important underlying electronic structure and phase relations, affecting both thermal conductivity of the material and its interaction with the cladding and coolant. In 2011, various XAS experiments on U1-xAmxO2±δ samples prepared by different synthesis methods have reported contradictory results on the charge distribution of U and Am. This work alleviates this discrepancy. The XAS results confirm that, independently of the synthesis process, the reductive sintering of U1-xAmxO2±δ leads to the formation of similar fluorite solid solution indicating the presence of Am+III and U+V in equimolar proportions.

  6. Study of interaction of uranium, plutonium and rare earth fluorides with some metal oxides in fluoric salt melts

    International Nuclear Information System (INIS)

    Interaction of plutonium, uranium, and rare-earth elements (REE) fluorides with aluminium and calcium oxides in melts of eutectic mixture LiF-NaF has been studied at 800 deg C by X-ray diffraction method. It has been shown that tetravalent uranium and plutonium are coprecipitated by oxides as a solid solution UO2-PuO2. Trivalent plutonium in fluorides melts in not precipitated in the presence of tetravalent uranium which can be used for their separation. REE are precipitated from a salt melt by calcium oxide and are not precipitated by aluminium oxide. Thus, aluminium oxide in a selective precipitator for uranium and plutonium in presence of REE. Addition of aluminium fluoride retains trivalent plutonium and REE in a salt melt in presence of Ca and Al oxides. The mechanism of interacting plutonium and REE trifluorides with metal oxides in fluoride melts has been considered

  7. Adsorption of plutonium, neptunium, and uranium on stainless steel from nitric acid

    International Nuclear Information System (INIS)

    Adsorption measurements have been made for plutonium, neptunium, and uranium in various valency states and as colloids on 12Kh18N10T steel at concentrations of 10-7-10-1 M. No matter what the state, the uptake increases with the concentration. The colloidal forms show the most adsorption, the levels for ionic ones being lower by an order of magnitude. Empirical equations are derived to relate the uptake to the concentration. The sorbent material affects the plutonium uptake

  8. Analysis of Enriched Uranium and Weapons Plutonium Reloads for PWRs Using BRACC

    International Nuclear Information System (INIS)

    Comparisons of the multicycle results demonstrate that the correlation coefficients based on the CASMO3 data were implemented correctly and that the Linear Reactivity Model is acceptably accurate for missed reloads containing both uranium and weapons plutonium fuel. The expanded set of correlation coefficients make BRACC a useful tool for performing multi-cycle in-core fuel management studies of PWR cores containing weapons plutonium

  9. Alpha spectroscopic determination of plutonium and uranium in food, biological materials, and soils

    International Nuclear Information System (INIS)

    An alpha-spectrometric method for the plutonium determination which was tested in different samples is described in detail. In particular, this method is capable of determining the very low plutonium levels found in food at present, and allow recoveries of 85-95% of the tracer added. Inorganic samples, such as soil samples for example, can be analyzed by using an abbreviated modification of the method. The measuring preparations show a high degree of spectral purity. Uranium can be separated during the analytical procedure and, after purification, can also be determined alpha-spectrometrically. 90-100% of the uranium are recovered. (orig.)

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

    OpenAIRE

    Golubev V.K.

    2012-01-01

    Results on studying the spall fracture of uranium, plutonium and several their alloys under shock wave loading are presented in the paper. The problems of influence of initial temperature in a range of − 196 – 800∘C and loading time on the spall strength and failure character of uranium and two its alloys with molybdenum and both molybdenum and zirconium were studied. The results for plutonium and its alloy with gallium were obtained at a normal temperature and in a temperature range of 40–31...

  11. Ingot formation using uranium dendrites recovered by electrolysis in LiCl-KCl-PuCl{sub 3}-UCl{sub 3} melt

    Energy Technology Data Exchange (ETDEWEB)

    Mineo Fukushima; Akira Nakayoshi; Shinichi Kitawaki [Japan Atomic Energy Agency (JAEA), 4-33 Muramatsu Tokai-mura Naka-gun, Ibaraki, 319-1194 (Japan); Masaki Kurata; Noboru Yahagi [Central Research Institute of Electric Power Industry (CRIEPI), 2-11-1 Iwadokita Komae-shi, Tokyo, 201-8511 (Japan)

    2008-07-01

    Products on solid cathodes recovered by the metal pyrochemical processing were processed to obtain uranium ingot. Studies on process conditions of uranium formation, assay recovered uranium products and by-products and evaluation of mass balance were carried out. In these tests, it is confirmed that uranium ingots can be obtained with heating the products more than melting temperature of metal uranium under atmospheric pressure because adhered salt cover the uranium not to oxidize it during uranium cohering. Covered salt can be removed after ingot formation. Inside the ingot, there were a lump of uranium and dark brown colored dross was observed. Material balance of uranium is 77 {approx} 96%, that of plutonium is 71 {approx} 109%, and that of americium that is a volatile substance more than uranium and plutonium become 79 {approx} 119%. Volatilization of americium is very small under the condition of high temperature. (authors)

  12. Isotope ratio analysis of individual plutonium and uranium-plutonium mixed oxide particles by thermal ionization mass spectrometry with a continuous heating method

    International Nuclear Information System (INIS)

    Isotope ratio analysis of individual plutonium and uranium-plutonium mixed oxide (MOX) particles was performed using a combination of single particle transfer and thermal ionization mass spectrometry (TIMS) with a continuous heating method, namely the particles were measured without any chemical treatment. Accurate values were obtained for all isotope ratios of the individual particles prepared from standard solution. As a consequence, it is confirmed that the combination method is useful for isotope ratio analysis of individual plutonium and MOX particles. (author)

  13. Transportability Class of Americium in K Basin Sludge under Ambient and Hydrothermal Processing Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmitt, Bruce E.; Schmidt, Andrew J.

    2006-08-01

    This report establishes the technical bases for using a ''slow uptake'' instead of a ''moderate uptake'' transportability class for americium-241 (241Am) for the K Basin Sludge Treatment Project (STP) dose consequence analysis. Slow uptake classes are used for most uranium and plutonium oxides. A moderate uptake class has been used in prior STP analyses for 241Am based on the properties of separated 241Am and its associated oxide. However, when 241Am exists as an ingrown progeny (and as a small mass fraction) within plutonium mixtures, it is appropriate to assign transportability factors of the predominant plutonium mixtures (typically slow) to the Am241. It is argued that the transportability factor for 241Am in sludge likewise should be slow because it exists as a small mass fraction as the ingrown progeny within the uranium oxide in sludge. In this report, the transportability class assignment for 241Am is underpinned with radiochemical characterization data on K Basin sludge and with studies conducted with other irradiated fuel exposed to elevated temperatures and conditions similar to the STP. Key findings and conclusions from evaluation of the characterization data and published literature are summarized here. Plutonium and 241Am make up very small fractions of the uranium within the K Basin sludge matrix. Plutonium is present at about 1 atom per 500 atoms of uranium and 241Am at about 1 atom per 19000 of uranium. Plutonium and americium are found to remain with uranium in the solid phase in all of the {approx}60 samples taken and analyzed from various sources of K Basin sludge. The uranium-specific concentrations of plutonium and americium also remain approximately constant over a uranium concentration range (in the dry sludge solids) from 0.2 to 94 wt%, a factor of {approx}460. This invariability demonstrates that 241Am does not partition from the uranium or plutonium fraction for any characterized sludge matrix. Most

  14. Kinetic study of the fluorination by fluorine of some uranium and plutonium compounds

    International Nuclear Information System (INIS)

    The study of fluorination reactions of uranium and plutonium compounds with elementary fluorine, has been carried out using a thermogravimetric method. These reactions are heterogeneous ones, and of the following type: S(solid) + G1(gas) - G2(gas). The kinetics of these reactions correspond to a uniform attack of the entire surface of the sample. α: being the degree of completion of the reaction, k(rel): being the relative rate of penetration of the reaction interface, t: being the time, one have the relation: (1-α)1/3 = 1 - k(rel)*t. The mechanism of the reaction varies according to the nature of the compound: 1) with uranium tetrafluoride and plutonium tetrafluoride, the reaction proceeds in a single step; 2) with uranium oxides, the reaction proceeds in two steps, uranium oxyfluoride being the intermediate compound; 3) with plutonium oxide, the reaction proceeds in two steps, plutonium tetrafluoride being the intermediate compound; and 4) with uranium trichloride, the mechanism is complex: chlorine trifluoride is formed. (author)

  15. Criticality experiments with mixed plutonium and uranium nitrate solution at a plutonium fraction of 0.4 in slab geometry

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, B.A.; Keeton, S.C.

    1997-09-01

    R. C. Lloyd of PNL has completed and published a series of critical experiments with mixed plutonium- uranium nitrate solutions (Reference 1). This series of critical experiments was part of an extensive program jointly sponsored by the U. S. Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan and was carried out in the mid-1980`s. The experiments evaluated here (published as Report PNL-6327) were performed with mixed plutonium- uranium nitrate solution in a variable thickness slab tank with two 106.7 cm square sides and a width that could be varied from 7.6 to 22.8 cm. The objective of these experiments was to obtain experimental data to permit the validation of computer codes for criticality calculations and of cross-section data to minimize the uncertainties inherent therein, so that facility safety, efficiency, and reliability could be enhanced. The concentrations of the solution were about 105, 293, and 435 g(Pu+U)/liter with a ratio of plutonium to total heavy metal (plutonium plus uranium) of about 0. 40 for all eight experiments. Four measurements were made with a water reflector, and four with no reflector. Following the publication of the initial PNL reports, considerable effort was devoted to an extensive reevaluation of this series of experiments by a collaboration of researchers from ORNL, PNL, and PNC (Reference 2). Their work resulted in a more accurate description of the ``as built`` hardware configuration and the materials specifications. For the evaluations in this report, the data published in Reference 2 by Smolen et al. is selected to supersede the original PNL report. Eight experiments have been evaluated and seven (063, 064, 071, 072, 074, 075, and 076) provide benchmark criticality data. Experiment 073 could not achieve criticality within vessel height limitations.

  16. Human bones obtained from routine joint replacement surgery as a tool for studies of plutonium, americium and 90Sr body-burden in general public

    International Nuclear Information System (INIS)

    The paper presents a new sampling method for studying in-body radioactive contamination by bone-seeking radionuclides such as 90Sr, 239+240Pu, 238Pu, 241Am and selected gamma-emitters, in human bones. The presented results were obtained for samples retrieved from routine surgeries, namely knee or hip joints replacements with implants, performed on individuals from Southern Poland. This allowed to collect representative sets of general public samples. The applied analytical radiochemical procedure for bone matrix is described in details. Due to low concentrations of 238Pu the ratio of Pu isotopes which might be used for Pu source identification is obtained only as upper limits other then global fallout (for example Chernobyl) origin of Pu. Calculated concentrations of radioisotopes are comparable to the existing data from post-mortem studies on human bones retrieved from autopsy or exhumations. Human bones removed during knee or hip joint surgery provide a simple and ethical way for obtaining samples for plutonium, americium and 90Sr in-body contamination studies in general public. - Highlights: → Surgery for joint replacement as novel sampling method for studying in-body radioactive contamination. → Proposed way of sampling is not causing ethic doubts. → It is a convenient way of collecting human bone samples from global population. → The applied analytical radiochemical procedure for bone matrix is described in details. → The opposite patient age correlations trends were found for 90Sr (negative) and Pu, Am (positive).

  17. The estimation of reactions of hematopoietic systems of organisms to the effect, caused by americium and plutonium, of nuclear industry workers

    Energy Technology Data Exchange (ETDEWEB)

    Gasteva, G. N.; Ivanova, T. A.; Gordeeva, A. A.; Suvorova, L. A.; Molokanov, A. A.; Badine, I.

    2004-07-01

    Object of research are the workers having in an organism radioactive substance (Am-241 and Pu-239). The purpose of work was the estimation of reaction hemopoietic systems of an organism on influence of americium and plutonium at workers of the nuclear industry. At the surveyed contingent of persons the determined effects caused by total influence Am-241 and Pu-239 are ascertained; chronic radiation disease with development, besides diffusive a pneumoscleoris and a chronic toxic-chemical radiating bronchitis, reactions of system of blood, jet hepatopathy which frequency accrued with increase doses loadings and essentially did not depend on age. In peripheral blood on the foreground jet changes act: hyperglobulia, the tendency to neutrophilus leukocytosis, monocytosis, increase ESR, decrease (reduction ?/G of factor reflecting weight and processing of defeat bronchus and pulmonary of system. Stable downstroke in number thrombocytes and reticulocytes in peripheral blood, their direct dependence on a doze of an irradiation, reflect hypoplastic a background hemogenesis, caused by long influence incorporatedin a bone and a bone brain of radioactive substances. At cytologic research punctate a bone brain jet changes which are expressed in increase of functional activity erythro-and myelopoiesiscome to light and provide compensatory reaction of peripheral blood. At histologic research of a bone brain and a bone fabric attributes of development atrophic process which is expressed in reduction of volume parenchyma a bone brain (a fatty atrophy) and dysplasia to a bone fabric are observed.

  18. Human bones obtained from routine joint replacement surgery as a tool for studies of plutonium, americium and {sup 90}Sr body-burden in general public

    Energy Technology Data Exchange (ETDEWEB)

    Mietelski, Jerzy W., E-mail: jerzy.mietelski@ifj.edu.pl [Henryk Niewodniczanski Institute of Nuclear Physics, Polish Academy of Sciences, Radzikowskiego 152, 31-342 Cracow (Poland); Golec, Edward B. [Traumatology and Orthopaedic Clinic, 5th Military Clinical Hospital and Polyclinic, Independent Public Healthcare Facility, Wroclawska 1-3, 30-901 Cracow (Poland); Orthopaedic Rehabilitation Department, Chair of Clinical Rehabilitation, Faculty of Motor of the Bronislaw Czech' s Academy of Physical Education, Cracow (Poland); Department of Physical Therapy Basics, Faculty of Physical Therapy, Administration College, Bielsko-Biala (Poland); Tomankiewicz, Ewa [Henryk Niewodniczanski Institute of Nuclear Physics, Polish Academy of Sciences, Radzikowskiego 152, 31-342 Cracow (Poland); Golec, Joanna [Orthopaedic Rehabilitation Department, Chair of Clinical Rehabilitation, Faculty of Motor of the Bronislaw Czech' s Academy of Physical Education, Cracow (Poland); Physical Therapy Department, Institute of Physical Therapy, Faculty of Heath Science, Jagiellonian University, Medical College, Cracow (Poland); Nowak, Sebastian [Traumatology and Orthopaedic Clinic, 5th Military Clinical Hospital and Polyclinic, Independent Public Healthcare Facility, Wroclawska 1-3, 30-901 Cracow (Poland); Orthopaedic Rehabilitation Department, Chair of Clinical Rehabilitation, Faculty of Motor of the Bronislaw Czech' s Academy of Physical Education, Cracow (Poland); Szczygiel, Elzbieta [Physical Therapy Department, Institute of Physical Therapy, Faculty of Heath Science, Jagiellonian University, Medical College, Cracow (Poland); Brudecki, Kamil [Henryk Niewodniczanski Institute of Nuclear Physics, Polish Academy of Sciences, Radzikowskiego 152, 31-342 Cracow (Poland)

    2011-06-15

    The paper presents a new sampling method for studying in-body radioactive contamination by bone-seeking radionuclides such as {sup 90}Sr, {sup 239+240}Pu, {sup 238}Pu, {sup 241}Am and selected gamma-emitters, in human bones. The presented results were obtained for samples retrieved from routine surgeries, namely knee or hip joints replacements with implants, performed on individuals from Southern Poland. This allowed to collect representative sets of general public samples. The applied analytical radiochemical procedure for bone matrix is described in details. Due to low concentrations of {sup 238}Pu the ratio of Pu isotopes which might be used for Pu source identification is obtained only as upper limits other then global fallout (for example Chernobyl) origin of Pu. Calculated concentrations of radioisotopes are comparable to the existing data from post-mortem studies on human bones retrieved from autopsy or exhumations. Human bones removed during knee or hip joint surgery provide a simple and ethical way for obtaining samples for plutonium, americium and {sup 90}Sr in-body contamination studies in general public. - Highlights: > Surgery for joint replacement as novel sampling method for studying in-body radioactive contamination. > Proposed way of sampling is not causing ethic doubts. > It is a convenient way of collecting human bone samples from global population. > The applied analytical radiochemical procedure for bone matrix is described in details. > The opposite patient age correlations trends were found for 90Sr (negative) and Pu, Am (positive).

  19. Actinide and lanthanum accumulation by immobilized cells of a citrobacter sp. and application to the decontamination of solutions containing americium and plutonium

    International Nuclear Information System (INIS)

    Phosphatase-mediated metal bioaccumulation by a Citrobacter sp. underlies a bioprocess for the removal of heavy metals from solution, as cell-bound metal phosphate. Deposition of uranyl ion indicated a role in the biotechnological removal of americium and plutonium from wastes generated from the nuclear fuel cycle. Preliminary studies suggested a recalcitrance of tetravalent species of U(IV), Th(IV) and Zr(IV) and, by implication, Pu(IV), probably attributable to the stability of metal-ligand complexes in solution. Trials with the trivalent model, La(III), indicated probable bioaccumulation of Pu(III) and Am(III), which was confirmed by the removal of 241Am by cells immobilized in a cartridge incorporated into a flow supplemented with Am. Pu(V) and Pu(IV) wastes may be treatable via prior reduction to Pu(III), with simultaneous removal of the latter with the co-contaminant Am(III). An oxidative route, to Pu(VI), with desolubilization as HPuO2PO4 was also considered, but experiments using the analogous U(VI) (uranyl ion) demonstrated a greater efficiency of M(III) removal. Initial experiments utilized polyacrylamide gel-immobilized cells. 241Am removal also occurred with Citrobacter sp. immobilized as biofilm on reticulated foam supports, more amenable to large-scale processes

  20. Neutron spectra in thorium and depleted uranium-plutonium-loaded light water reactors

    International Nuclear Information System (INIS)

    The technical feasibility of using plutonium mixed with natural uranium in one-third of the cores of light water reactors (LWRs) has been sufficiently demonstrated. A number of reactors in Europe are currently operated with one-third mixed-oxide cores. If the option of burning excess plutonium in conventional LWR reactors in this country is selected, it has been estimated that the long-term disposition of the excess plutonium would take many decades. This time can be significantly reduced if the plutonium is burned in a fast breeder reactor. However, in the present economic and political climate, such an approach is difficult to implement. On the other hand, if the neutron spectrum in an LWR core is hardened, the well-developed and well-understood LWR can accomplish the goal of effectively burning excess plutonium to convert to proliferation-resistive fuel such as 233U. The authors present some fundamental characteristics of thorium and depleted uranium-plutonium-fueled LWRs. High fuel burnup levels can be achieved by tightening the lattice of an LWR loaded with thorium and depleted uranium and plutonium (nitride or oxide) and increasing the plutonium content. The neutron spectrum in such a reactor is very hard and tends to approach that of a Na-cooled fast reactor. Instead of Zircaloy, stainless steel can be used as fuel cladding and structural material since it has a low fast neutron capture cross section. Supercritical steam, generated at high pressures, can be used as coolant. If the cladding and structural materials used in this reactor can withstand corrosion in water under high-irradiation conditions, high conversion ratios of thermal heat to electricity will be possible

  1. Post irradiation examination of irradiated americium oxide and uranium dioxide in magnesium aluminate spinel

    International Nuclear Information System (INIS)

    To study MgAl2O4 spinel as inert matrix material for the transmutation of minor actinides, two capsules were irradiated at the high flux reactor in Petten, containing 12.5 wt% micro-dispersed 241AmOx in spinel and 25 wt% micro-dispersed enriched UO2 in spinel. During irradiation, the initially present 241Am was converted for 99.8% to fission products (50%), plutonium (30%), curium (16%) and 243Am (4%). The UO2 spinel target experienced a burn-up of 32% fission per initial metal atom. The post irradiation examination of the AmOx inert matrix target showed swelling of 27 vol.%, and a gas release of 48% for He and 16% for Xe and Kr. The UO2 inert matrix target also showed a large volumetric swelling of 11%, directed mainly radially. Ceramography on the UO2 inert matrix target revealed a complete restructuring of the spinel grains upon irradiation and the absence of porosity, suggesting that amorphisation is the main cause of the swelling

  2. Post irradiation examination of irradiated americium oxide and uranium dioxide in magnesium aluminate spinel

    Science.gov (United States)

    Klaassen, F. C.; Bakker, K.; Schram, R. P. C.; Klein Meulekamp, R.; Conrad, R.; Somers, J.; Konings, R. J. M.

    2003-06-01

    To study MgAl 2O 4 spinel as inert matrix material for the transmutation of minor actinides, two capsules were irradiated at the high flux reactor in Petten, containing 12.5 wt% micro-dispersed 241AmO x in spinel and 25 wt% micro-dispersed enriched UO 2 in spinel. During irradiation, the initially present 241Am was converted for 99.8% to fission products (50%), plutonium (30%), curium (16%) and 243Am (4%). The UO 2 spinel target experienced a burn-up of 32% fission per initial metal atom. The post irradiation examination of the AmO x inert matrix target showed swelling of 27 vol.%, and a gas release of 48% for He and 16% for Xe and Kr. The UO 2 inert matrix target also showed a large volumetric swelling of 11%, directed mainly radially. Ceramography on the UO 2 inert matrix target revealed a complete restructuring of the spinel grains upon irradiation and the absence of porosity, suggesting that amorphisation is the main cause of the swelling.

  3. Post irradiation examination of irradiated americium oxide and uranium dioxide in magnesium aluminate spinel

    Energy Technology Data Exchange (ETDEWEB)

    Klaassen, F.C. E-mail: klaassen@nrg-nl.com; Bakker, K.; Schram, R.P.C.; Klein Meulekamp, R.; Conrad, R.; Somers, J.; Konings, R.J.M

    2003-06-01

    To study MgAl{sub 2}O{sub 4} spinel as inert matrix material for the transmutation of minor actinides, two capsules were irradiated at the high flux reactor in Petten, containing 12.5 wt% micro-dispersed {sup 241}AmO{sub x} in spinel and 25 wt% micro-dispersed enriched UO{sub 2} in spinel. During irradiation, the initially present {sup 241}Am was converted for 99.8% to fission products (50%), plutonium (30%), curium (16%) and {sup 243}Am (4%). The UO{sub 2} spinel target experienced a burn-up of 32% fission per initial metal atom. The post irradiation examination of the AmO{sub x} inert matrix target showed swelling of 27 vol.%, and a gas release of 48% for He and 16% for Xe and Kr. The UO{sub 2} inert matrix target also showed a large volumetric swelling of 11%, directed mainly radially. Ceramography on the UO{sub 2} inert matrix target revealed a complete restructuring of the spinel grains upon irradiation and the absence of porosity, suggesting that amorphisation is the main cause of the swelling.

  4. Gender-specific differences in the biokinetics of plutonium and other actinides

    International Nuclear Information System (INIS)

    The published data on the biokinetics of actinide distribution and retention in humans and animals have been reviewed. It is concluded that in humans there are strong indications that urinary and faecal excretion of plutonium is greater in females than males. Animal data indicate that this may also be true for americium, neptunium and uranium. (author)

  5. Reduction of uranium and plutonium oxides by aluminum. Application to the recycling of plutonium; Reduction des oxydes d'uranium et de plutonium par l'aluminium application au recyclage du plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Gallay, J. [Commissariat a l' Energie Atomique, Valduc (France). Centre d' Etudes

    1968-07-01

    A process for treating plutonium oxide calcined at high temperatures (1000 to 2000 deg. C) with a view to recovering the metal consists in the reduction of this oxide dissolved in a mixture of aluminium, sodium and calcium fluorides by aluminium at about 1180 deg. C. The first part of the report presents the results of reduction tests carried out on the uranium oxides UO{sub 2} and U{sub 3}O{sub 8}; these are in agreement with the thermodynamic calculations of the exchange reaction at equilibrium. The second part describes the application of this method to plutonium oxides. The Pu-Al alloy obtained (60 per cent Pu) is then recycled in an aqueous medium. (author) [French] Un procede de traitement de l'oxyde de plutonium calcine a haute temperature (1000 deg. C a 2000 deg. C), en vue de la recuperation du metal, consiste a reduire cet oxyde dissous dans un melange de fluorures d'aluminium, de sodium et de calcium, par l'aluminium vers 1180 deg. C. Une premiere partie du rapport presente les resultats des essais de reduction des oxydes d'uranium UO{sub 2} et U{sub 3}O{sub 8}, en accord avec les resultats du calcul thermodynamique de la reaction d'echange a l'equilibre. Une seconde partie rend compte de l'application de cette methode a l'oxyde de plutonium. L'alliage Pu-Al obtenu (60 pour cent Pu) est ensuite recycle par voie aqueuse. (auteur)

  6. Separation and determination of americium in low-level alkaline waste of NPP origin

    Science.gov (United States)

    Todorov, B.; Djingova, R.; Nikiforova, A.

    2006-01-01

    The aim of this work is to develop a short and cost-saving procedure for the determination of 241Am in sludge sample of the alkaline low-level radioactive waste (LL LRAW) collected from Nuclear Power Plant “Kozloduy”. The determination of americium was a part of a complex analytical approach, where group actinide separation was achieved. An anion exchange was used for separation of americium from uranium, plutonium and iron. For the separation of americium extraction with diethylhexyl phosphoric acid (DEHPA) was studied. The final radioactive samples were prepared by micro co-precipitation with NdF3, counted by alpha and gamma spectrometry. The procedure takes 2 hours. The recovery yield of the procedure amounts to (95 ± 1.5)% and the detection limit is 53 mBq/kg 241Am (t=150 000 s). The analytical procedure was applied for actual liquid wastes and results were compared to standard procedure.

  7. An intake prior for the bayesian analysis of plutonium and uranium exposures in an epidemiology study

    International Nuclear Information System (INIS)

    In Bayesian inference, the initial knowledge regarding the value of a parameter, before additional data are considered, is represented as a prior probability distribution. This paper describes the derivation of a prior distribution of intake that was used for the Bayesian analysis of plutonium and uranium worker doses in a recent epidemiology study. The chosen distribution is log- normal with a geometric standard deviation of 6 and a median value that is derived for each worker based on the duration of the work history and the number of reported acute intakes. The median value is a function of the work history and a constant related to activity in air concentration, M, which is derived separately for uranium and plutonium. The value of M is based primarily on measurements of plutonium and uranium in air derived from historical personal air sampler (PAS) data. However, there is significant uncertainty on the value of M that results from paucity of PAS data and from extrapolating these measurements to actual intakes. This paper compares posterior and prior distributions of intake and investigates the sensitivity of the Bayesian analyses to the assumed value of M. It is found that varying M by a factor of 10 results in a much smaller factor of 2 variation in mean intake and lung dose for both plutonium and uranium. It is concluded that if a log-normal distribution is considered to adequately represent worker intakes, then the Bayesian posterior distribution of dose is relatively insensitive to the value assumed of M. (authors)

  8. Determination of plutonium and uranium in mixed oxide fuels by sequential redox titration.

    Science.gov (United States)

    Chadwick, P H; McGowan, I R

    1972-11-01

    The use of a sequential determination of uranium and plutonium in a single sample solution results in a saving in analysis time and apparatus requirements. The method starts with U(IV) and Pu(in) in a mixture of sulphuric and nitric adds. Titration with dichromate, using amperometry at a pair of polarizable electrodes, produces two well-defined end-points corresponding to the sequential oxidation of U(IV) to U(VI) and Pu(III) to Pu(IV). The quantitative oxidation of U(IV) to U(VI) is achieved via the action of Pu(IV) as intermediate, and is dependent upon establishing conditions which favour rapid reaction between U(IV) and Pu(IV). The method is precise and accurate. With Pu-U mixtures containing between 15 and 30% plutonium the precision (3sigma) of the Pu: U ratio results is +/-0.6% on samples containing 100-120 mg of plutonium plus uranium. Iron and vanadium interfere quantitatively with plutonium, copper interferes non-quantitatively with uranium, and gross amounts of molybdenum mask the uranium end-point. PMID:18961189

  9. Irradiation of the experimental fuel assemblies with uranium-plutonium fuel in the BN-600 reactor

    International Nuclear Information System (INIS)

    Design features of experimental fuel assemblies (EFA) with uranium-plutonium mixed oxide fuel specific aspects of their arrangement within the BN-600 reactor core, conditions and basic results of EFA with the fuel mentioned in the BN-600 reactor are described

  10. Reactions of plutonium and uranium with water: Kinetics and potential hazards

    International Nuclear Information System (INIS)

    The chemistry and kinetics of reactions between water and the metals and hydrides of plutonium and uranium are described in an effort to consolidate information for assessing potential hazards associated with handling and storage. New experimental results and data from literature sources are presented. Kinetic dependencies on pH, salt concentration, temperature and other parameters are reviewed. Corrosion reactions of the metals in near-neutral solutions produce a fine hydridic powder plus hydrogen. The corrosion rate for plutonium in sea water is a thousand-fold faster than for the metal in distilled water and more than a thousand-fold faster than for uranium in sea water. Reaction rates for immersed hydrides of plutonium and uranium are comparable and slower than the corrosion rates for the respective metals. However, uranium trihydride is reported to react violently if a quantity greater than twenty-five grams is rapidly immersed in water. The possibility of a similar autothermic reaction for large quantities of plutonium hydride cannot be excluded. In addition to producing hydrogen, corrosion reactions convert the massive metals into material forms that are readily suspended in water and that are aerosolizable and potentially pyrophoric when dry. Potential hazards associated with criticality, environmental dispersal, spontaneous ignition and explosive gas mixtures are outlined

  11. Preparation of uranium-plutonium mixed nitride pellets with high purity

    International Nuclear Information System (INIS)

    Uranium-plutonium mixed nitride pellets have been prepared in the gloveboxes with high purity Ar gas atmosphere. Carbothermic reduction of the oxides in N2-H2 mixed gas stream was adopted for synthesizing mixed nitride. Sintering was carried out in various conditions and the effect on the pellet characteristics was investigated. (author)

  12. Equation of state and transport properties of uranium and plutonium nitrides in the liquid region

    International Nuclear Information System (INIS)

    By the use of available low-temperature data for various thermophysical and transport properties for uranium and plutonium nitrides, values above the melting point of density, heat capacity, enthalpy, vapor pressure, thermal conductivity, and viscosity were estimated. Sets of recommended values have been prepared for the compounds UN, PuN, and (U0.8Pu0.2)N

  13. Analysis of trace uranium and plutonium in environmental water sample by ICP-MS

    International Nuclear Information System (INIS)

    The analysis of trace Uranium and Plutonium in environmental water is very important in the environment inspect. The preparation method of water samples are introduced and several common used method are compared. The analysis process and the calibration method with ICP-MS are discussed in detail considering present conditions. (author)

  14. Simulation study for purification, recovery of plutonium and uranium from plant streams of Fast Reactor Fuel Reprocessing Plant

    International Nuclear Information System (INIS)

    A method for removal of plutonium from the lean organic streams obtained after co-stripping of uranium -plutonium was developed. Plutonium from lean organic phase was stripped using U4+/hydrazine as the stripping agent. The effect of concentrations of stripping agent U4+ and feed Pu concentration in the lean organic phase was studied. Lean organic phases having higher plutonium concentration require three stages of stripping to bring plutonium concentration 4+ stabilized by hydrazine reduces Pu (IV) to Pu (III) thereby stripping plutonium from the organic phase. The non-extractability of Pu (III) by TBP was utilized for development of flow sheet for obtaining a uranium product lean of plutonium for ease of handling. (author)

  15. Extraction of uranium(VI) and plutonium(IV) with unsymmetrical monoamides

    International Nuclear Information System (INIS)

    Solvent extraction of uranium(VI) and plutonium(IV) has been investigated by three unsymmetrical monoamides, viz., N-methyl, N-butyl derivatives of hexanamide (MBHA), octanamide (MBOA) and decanamide (MBDA) in n-dodecane at 25, 30, 40 and 50 C (±0.1). The extraction of the metal ions follows the order of the basicity of the amides (except for plutonium (IV)-MBDA system) as indicated by the equilibrium constant for the uptake of nitric acid by the amide (KH). The order of the basicity was found to increase with the length of the carbon chain, viz., MBHA< MBOA< MBDA. Uranium(VI) and plutonium(IV) were found to be extracted as disolvates and trisolvates respectively, as was observed in the case of symmetrical monoamides. The thermodynamic parameters determined by the temperature coefficient method indicate the extraction reactions to be predominantly enthalpy stabilized. (orig.)

  16. Simultaneous determination of uranium and plutonium at trace levels in process streams using Arsenazo III by derivative spectrophotometry

    International Nuclear Information System (INIS)

    A derivative spectrophotometric method has been developed for the simultaneous determination of uranium and plutonium at trace levels in various process streams in 3M HNO3 medium using Arsenazo III. The method was developed with the objective of measuring both uranium and plutonium in the same aliquot in fairly high burn-up fuels. The first derivative absorbances of the uranium and plutonium Arsenazo III complexes at 632 nm and 606.5 nm, respectively, were used for their quantification. Mixed aliquots of uranium (20-28 μg/ml) and plutonium (0.5-1.5 μg/ml) with U/Pu ratio varying from 25 to 40 were analysed using this technique. A relative error of about 5% was obtained for uranium and plutonium. The method is simple, fast and does not require separation of uranium and plutonium. The effect of presence of many fission products, corrosion products and complexing anions on determination of uranium and plutonium was also studied. (author)

  17. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    International Nuclear Information System (INIS)

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% totalPu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products

  18. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  19. Uranium utilization in the closed HWR cycle assuming self-generated recycling of plutonium

    International Nuclear Information System (INIS)

    In order to illustrate the case of single independent reactor, in this report the Self-Generated Recycle Mode is considered, where only plutonium generated in the previous cycles is recycled. The topping material is natural uranium. A simple mode of transition from the low burnup natural uranium UO2 fuel of the initial core to the higher burnup MOX fuel was selected, which leads to a bundle-to-bundle variable out-of-core delay time for the fuel elements of the initial and first generation cores. This delay time increases continuously so that: MOX fuel out-of-core delay time/initial nat.uranium UO2 fuel out-of-core delay time = MOX fuel burnup/equil.nat.uranium UO2 fuel burnup. It is concluded that the recycling of self-generated plutonium results in natural uranium savings of about 53% relative to the natural uranium once through cycle. The fissile Pu inventory is about 10 times less than in the stored spent fuel elements of the natural uranium once-through cycle

  20. Removal of plutonium from low level Purex waste streams by Chitosan

    International Nuclear Information System (INIS)

    The low level waste solution generated from Purex process contains traces of Plutonium and Americium contributing alpha activity to the solution. Chitosan is it natural bio-polymer derived from Chitin. Successful studies were carried out using Chitosan to recover the uranium, thorium and americium from different waste streams. The studies were extended to find out its plutonium sorption characteristics. Chitosan was equilibrated with pure plutonium tracer solution at different pH, for 60 minutes with a Chitosan to aqueous ratio of 1:100 and the raffinates were filtered and analysed radio metrically. The results showed -95 % of plutonium could be recovered by Chitosan between pH 4 and 7. Elution study of loaded plutonium was also studied with 1M HNO3. (author)

  1. Evaluation of anion exchange resins for plutonium-uranium separations in nitric acid

    International Nuclear Information System (INIS)

    Pellicular, macroreticular and microreticular (gel type) anion exchange resins were compared for separation of plutonium from nitric acid solutions of mixed plutonium-uranium. All the macroreticular resins were 20 to 50 mesh beads. Dowex 1-X4 gel resin was 50 to 80 mesh beads. The resins were held in glass columns with coarse glass frits at the bottom of the columns. The top of the columns contained 50 ml reservoirs. The flow rates were controlled at 4 cm3.min-1.cm-2. One-centimeter bore columns with 15-cm resin bed heights were used for the plutonium elution and breakthrough capacity experiments, whereas 1.7 cm bore columns with 20 cm bed heights were used for the uranium washing experiments. As Pellionex SAX (pellicular resin) and Amberlite IRA-93 (weak base macroreticular anion exchange resin) were found to have better uranium washing and plutonium eluating characteristics than any of the resins tested. However, the capacity of the pellicular resin was much lower than that of the other resins. (T.G.)

  2. Plutonium waste crib logging using the prompt fission neutron uranium logging system

    International Nuclear Information System (INIS)

    Sandia Laboratories' Uranium Logging Project has demonstrated their prompt fission neutron (PFN) logging system at the Hanford, WA, site for Rockwell-Hanford Operations (RHO). The dates of the demonstration were July 31 through August 2, 1979. The purpose was to show RHO the capabilities of the system for measuring plutonium concentration. An underground effluent disposal crib associated with their processing facilities was used as the test site. The performance criterion was to be able to detect a 10 nCi/g concentration of plutonium. Six test wells penetrating the crib were logged, as were three other wells. The PFN tool was able to maintain a good signal-to-noise ratio even under the most extreme conditions of high count rate and high background. The wells at the center of the crib indicated very high concentrations of plutonium, while those at the periphery indicated much less. Concentrations estimated to be lower than 10 nCi/g were detected. Comparisons with core data were not made. The technique used to obtain physical samples for analysis did not follow uranium-exploration coring practice so comparisons were not possible. The data interpretation model used was originally developed for uranium and was modified to calculate plutonium concentration. Results indicated that the operation of a PFN logging system by RHO personnel would provide a suitable technique for monitoring transuranic waste storage sites

  3. Preparation of the first electrolytic plutonium and of uranium from fused chlorides

    International Nuclear Information System (INIS)

    The production at Los Alamos of the first electrolytic plutonium is described. Because it was scarce, it was initially prepared on a 50 mg scale. The electrolyte consisted of PuCl3 dissolved in a ternary eutectic o BaCl2-KCl-NaCl and was operated at 6600C. The technique involved the electrolysis of as little as 0.1 cm3 of fused salt in a glass cell to produce liquid metal. Subsequently, a few successful 1g electrolyses were carried out with a maximum recovery of 75% of the plutonium content of the bath. It was found that the plutonium trichloride had to be dry and relatively low in oxygen and was, therefore, best prepared by dry methods. Electrolyses were performed under purified argon or hydrogen using tungsten or graphite anodes. The metallic product was pure and was used for early development of methods for cleaning, etching, and plating plutonium. Uranium was used for finding and testing techniques before plutonium became available. Acceptable deposits could not be obtained from UCl4 dissolved in the ternary chloride eutectic, but were produced when UCl3 was substituted for the tetrachloride. At about 6500C, the uranium was deposited in the form of solid, dendriti crystals. These were, in general purer than the starting materials and were readily converted to massive metal

  4. Equipment for manufacture of uranium-plutonium mixed carbide fuel pins

    International Nuclear Information System (INIS)

    The equipment for manufacturing fuel pins, which is neccesary for irradiation tests of uranium-plutonium mixed carbide fuels, has been provided. This equipment is composed of a centerless grinder, an apparatus for loading fuel pellets and endplugs into cladding tubes, a TIG-welder, an apparatus for decontamination of welded fuel pins, and so on. Most of them are installed into gloveboxes, in order to prevent the workers from plutonium contamination. The maximum size of the pins manufactured by the equipment is 15 mm in radius and 600 mm in length. In this report, design, construction, and ability of the equipment for the manufacture of carbide fuel pins are described. (author)

  5. Nonproliferation analysis of the reduction of excess separated plutonium and high-enriched uranium

    International Nuclear Information System (INIS)

    The purpose of this preliminary investigation is to explore alternatives and strategies aimed at the gradual reduction of the excess inventories of separated plutonium and high-enriched uranium (HEU) in the civilian nuclear power industry. The study attempts to establish a technical and economic basis to assist in the formation of alternative approaches consistent with nonproliferation and safeguards concerns. The analysis addresses several options in reducing the excess separated plutonium and HEU, and the consequences on nonproliferation and safeguards policy assessments resulting from the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials

  6. Radiotoxicity and decay heat power of spent uranium-plutonium and thorium fuel at long-term storage

    International Nuclear Information System (INIS)

    Changes of radiotoxicity and decay heat power of actinides from spent uranium- plutonium and thorium nuclear fuel of WWER-1000 type reactors at storage during 300 years are investigated in report. (author)

  7. Power and openness of 'Cogema'. Management of uranium and plutonium

    International Nuclear Information System (INIS)

    In the paper the 'Cogema' group activity in all stages of nuclear industrial cycle is covered. It is noticed, that 'Cogema' have joint ventures in the field of uranium wells development in the different countries of the world. In March of 1996 'Cogema' jointly with the National Atomic Company 'Kazatomprom' (Kazakhstan) the 'Katko' joint venture have implemented. J V 'Katko' posses with two licences on uranium ores mining for a 25 year term. Use of 'Muyunkum' uranium deposit (South Kazakhstan) carrying out by the mean of leaching technology with following ores reprocessing at the pilot plant. Capacity of the plant is 100 t of commercial uranium concentrate production per year. To middle of the summer of 2001 the plant was put into operation

  8. Simulation and analysis of experimental plutonium production rates in depleted uranium sphere bombarded by 14 MeV neutron

    International Nuclear Information System (INIS)

    According to the measurement of plutonium production rates resulting from 14 MeV neutrons in depleted uranium sphere, the experimental results and influencing factors were simulated and analyzed with MCNP5 code. It shows that simulant plutonium production rates from uresa and endf66c agree very well with experimental results, simulant effects on plutonium production rates by cavity, construction materials and water lays and background neutrons etc., are slight. (authors)

  9. Effects of Hanford high-level waste components on the solubility and sorption of cobalt, strontium, neptunium, plutonium, and americium

    International Nuclear Information System (INIS)

    High-level radioactive defense waste solutions, originating from plutonium recovery and waste processing operations at the U.S. Department of Energy's Hanford Site, currently are stored in mild steel-lined concrete tanks located in thick sedimentary beds of sand and gravel. Statistically designed experiments were used to identify the effects of 12 major chemical components of Hanford waste solution on radionuclide solubility and sorption. The chemical components with the most effect on radioelement solubility and sorption were NaOH, NaAl0/sub 2/, EDTA, and HEDTA. The EDTA and HEDTA increased Co, Sr, and Am solubility and decreased sorption for almost all radioelements studied. Sodium hydroxide and NaAlO/sub 2/ increased Pu solubility and decreased Np and Pu sorption. Sodium nitride decreased Np solubility, while Na/sub 2/CO/sub 3/ and HEDTA increased it. These observations give evidence for the formation of radioelement complexes which are soluble and are not strongly sorbed by the sediments near the waste tanks

  10. Flow sheet development studies on partitioning of uranium and plutonium during reprocessing of fast reactor fuels. Contributed Paper PE-07

    International Nuclear Information System (INIS)

    Separation of uranium and plutonium after achieving the sufficient decontamination factor from fission product is essential. An aqueous phase based process by the addition of uranous followed by solvent extraction with 30 % TBP is optimized. The results of this study indicate it is feasible to achieve the uranium and plutonium products such they can be reconverted in the respective oxides in reconversion laboratory. The results also indicates that the requirement of uranous is 10% more than the stochiometric requirement. (author)

  11. Experience on determination trace amount of uranium in plutonium oxide samples by ID-TIMS technique using indigenous TIMS (III)

    International Nuclear Information System (INIS)

    ID- TIMS method using 233U as spike and an alternative procedure adopted for ion exchange separation of uranium from bulk of Plutonium in Plutonium oxide samples was accurate and precise. The acetic acid anion exchange procedure has a better separation than compared to the standard ion exchange procedure, which was observed by the accuracy and precision values obtained in our experiments

  12. Effects of periodic manual stirring and uranium addition on surrogate plutonium glass processing

    International Nuclear Information System (INIS)

    Using thorium as a plutonium surrogate, homogeneous glasses have been processed ranging from 15 to 20 elemental weight percent thorium with 0 to 5 elemental weight percent uranium. Homogeneous glasses have been processed at 1,475 C with residence times ranging from 1 to 12 hours. High ramp rates successfully inhibited thorium silicate formation. Residence times of 5 to 7 hours were required for static melts to become homogeneous for glasses containing 15 elemental weight percent thorium. Thorium dissolution rates have been determined for glasses containing 15 elemental weight percent thorium with and without the addition of uranium. When compared to identical glass compositions which were not stirred, stirred melts produced homogeneous vitreous products with an 80 percent reduction in residence time. A 20 elemental weight percent thorium glass was produced at 1,475 C by adding 2 elemental weight percent uranium. Without the addition of uranium, a melt temperature of 1,500 C was required

  13. Disposition of PUREX facility tanks D5 and E6 uranium and plutonium solutions. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Harty, D.P.

    1993-12-01

    Approximately 9 kilograms of plutonium and 5 metric tons of uranium in a 1 molar nitric acid solution are being stored in two PUREX facility vessels, tanks D5 and E6. The plutonium was accumulated during cleanup activities of the plutonium product area of the PUREX facility. Personnel at PUREX recently completed a formal presentation to the Surplus Materials Peer Panel (SMPP) regarding disposition of the material currently in these tanks. The peer panel is a group of complex-wide experts who have been chartered by EM-64 (Office of Site and Facility Transfer) to provide a third party independent review of disposition decisions. The information presented to the peer panel is provided in the first section of this report. The panel was generally receptive to the information provided at that time and the recommendations which were identified.

  14. Disposition of PUREX facility tanks D5 and E6 uranium and plutonium solutions

    International Nuclear Information System (INIS)

    Approximately 9 kilograms of plutonium and 5 metric tons of uranium in a 1 molar nitric acid solution are being stored in two PUREX facility vessels, tanks D5 and E6. The plutonium was accumulated during cleanup activities of the plutonium product area of the PUREX facility. Personnel at PUREX recently completed a formal presentation to the Surplus Materials Peer Panel (SMPP) regarding disposition of the material currently in these tanks. The peer panel is a group of complex-wide experts who have been chartered by EM-64 (Office of Site and Facility Transfer) to provide a third party independent review of disposition decisions. The information presented to the peer panel is provided in the first section of this report. The panel was generally receptive to the information provided at that time and the recommendations which were identified

  15. Recycling of plutonium and uranium in water reactor fuel. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The Technical Committee Meeting on Recycling of Plutonium and Uranium in Water Reactor Fuel was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its aim was to obtain an overall picture of MOX fabrication capacity and technology, actual performance of this kind of fuel, and ways explored to dispose of the weapons grade plutonium. The subject of this meeting had been reviewed by the International Atomic Energy Agency every 5 to 6 years and for the first time the problem of weapons grade plutonium disposal was included. The papers presented provide a summary of experience on MOX fuel and ongoing research in this field in the participating countries. The meeting was hosted by British Nuclear Fuels plc, at Newby Bridge, United Kingdom, from 3 to 7 July 1995. Fifty-six participants from twelve countries or international organizations took part. Refs, figs, tabs

  16. Advances on reverse strike co-precipitation method of uranium-plutonium mixed solutions

    International Nuclear Information System (INIS)

    The reverse strike coprecipitation of uranium-plutonium mixed solutions, is an alternative way to obtain MOX fuel pellets. Previous tests, carried out in the Alpha Laboratory, included a stabilization step for transforming 100 % of plutonium into Pu+4 . Therefore, the plutonium precipitated as Pu(OH)4. In this second step, the stabilization process was suppressed. In this way, besides Pu(OH)4, a part of the precipitated is composed of a mixed salt: AD(U,Pu). Then, a homogeneous solid solution is formed in the early steps of the process. The powders showed higher tap density, better performance during the pressing and lower sinterability than the powders obtained in previous tests. The advantageous and disadvantageous effects of the stabilization step are analyzed in this paper. (author)

  17. Toxicity of uranium and plutonium to the developing embryos of fish

    International Nuclear Information System (INIS)

    The radiological and chemical toxicity of plutonium and uranium to the developing embryos of fish was investigated using eggs from carp, Cyprinus carpio, and fathead minnows, Pimephales promelas. Freshly fertilized eggs were developed in solutions containing high specific activity 238Pu or 232U or low specific activity 244Pu, 235U, or 238U. Quantitative tests to determine the penetration of these elements through the chorion indicated that plutonium accumulated in the contents of carp eggs reaching a maximum concentration factor of approximately 3.0 at hatching. Autoradiographs of 16 μ egg sections showed that plutonium was uniformly distributed in the egg volume. Uranium localized in the yolk material, and the concentration factor in the yolk sac remained constant during development at approximately 3.3. Doses from 238Pu which affected hatchability of the eggs were estimated to be 1.6 x 104 rads and 9.7 x 103 rads for C. carpio and P. promelas, respectively; doses from 232U were 1.3 x 104 rads for C. carpio and 2.7 x 103 rads for P. promelas. A greater number of abnormal larvae than in control groups was produced by 238Pu doses of 4.3 x 103 rads to carp and 5.7 x 102 rads to fathead minnows; 3.2 x 103 rads and 2.7 x 102 rads were estimated from 232U. Eggs that were incubated in 20 ppM 244Pu did not hatch. This mortality may have been the result of chemical toxicity of plutonium. Concentrations of 60 ppM of 235U and 238U did not affect egg hatching. Based on these data, concentrations in fish eggs were calculated for representative concentrations of uranium and plutonium in natural waters and the corresponding dose levels are below those levels at which observable effects begin to occur

  18. Computer programs for data reduction and interpretation in plutonium and uranium analysis by gamma ray spectrometry

    International Nuclear Information System (INIS)

    Non destructive gamma ray have been developed for analysis of isotopic abundances and concentrations of plutonium and uranium in the respective product solutions of a reprocessing plant. The method involves analysis of gamma rays emitted from the sample and uses a multichannel analyser system. Data reduction and interpretation of these techniques are tedious and time consuming. In order to make it possible to use them in routine analysis, computer programs have been developed in HP-BASIC language which can be used in HP-9845B desktop computer. A set of programs, for plutonium estimation by high resolution gamma ray spectrometry and for on-line measurement of uranium by gamma ray spectrometry are described in this report. (author) 4 refs., 3 tabs., 6 figs

  19. Presence of uranium and plutonium in marine sediments from gulf of Tehuantepec, Mexico

    International Nuclear Information System (INIS)

    Uranium and plutonium were determined in the Tehua II-21 sediment core collected from the Gulf of Tehuantepec, Mexico. The analyses were performed using radiochemical separation and alpha spectroscopy. Activity concentrations of alpha emitters in the sediment samples were from 2.56 to 43.1 Bq/kg for 238U, from 3.15 to 43.1 Bq/kg for 234U and from 0.69 to 2.95 Bq/Kg for 239+240Pu. Uranium activity concentration in marine sediment studied is generally high compared with those found in sediments from other marine coastal areas in the world. The presence of relatively high concentrations of anthropogenic plutonium in the sediments from the Gulf of Tehuantepec suggests that anthropogenic radionuclides have been incorporated and dispersed into the global marine environment. (author)

  20. Computational study of tetravalent uranium and plutonium lattice diffusion in zircon

    International Nuclear Information System (INIS)

    Empirical potentials have been established for zircon (ZrSiO4), uranium dioxide (UO2) and plutonium dioxide (PuO2) with the pair interactions U-O and Pu-O being transferable to zircon. The quality of the potentials obtained is tested by calculating different physical properties of these oxides and comparing to the experimental values. The transferability to zircon of the two body short range interactions, U-O and Pu-O, is tested by calculating the cell volume variation of orthosilicates ASiO4 (A=Zr,U,Pu) with respect to the contained A4+ ionic radius. Using the established force field and applying static transition state theory, we calculate the activation energies for lattice diffusion of uranium and plutonium in zircon. The corresponding diffusion coefficients are estimated and compared to recent experimental data. (orig.)

  1. Effect of cooling rate on achieving thermodynamic equilibrium in uranium-plutonium mixed oxides

    Science.gov (United States)

    Vauchy, Romain; Belin, Renaud C.; Robisson, Anne-Charlotte; Hodaj, Fiqiri

    2016-02-01

    In situ X-ray diffraction was used to study the structural changes occurring in uranium-plutonium mixed oxides U1-yPuyO2-x with y = 0.15; 0.28 and 0.45 during cooling from 1773 K to room-temperature under He + 5% H2 atmosphere. We compare the fastest and slowest cooling rates allowed by our apparatus i.e. 2 K s-1 and 0.005 K s-1, respectively. The promptly cooled samples evidenced a phase separation whereas samples cooled slowly did not due to their complete oxidation in contact with the atmosphere during cooling. Besides the composition of the annealing gas mixture, the cooling rate plays a major role on the control of the Oxygen/Metal ratio (O/M) and then on the crystallographic properties of the U1-yPuyO2-x uranium-plutonium mixed oxides.

  2. Determination of Uranium plus Plutonium by Alpha spectrometry in different matrix

    International Nuclear Information System (INIS)

    Usually, the determination of alpha emitters by alpha spectrometry is performed with a prior purification of each of the elements to be quantified. In this work, a methodology for the determination of uranium and plutonium isotopes as jointly described, in order to improve analytical processing times and measurement. The method includes purifying uranium and plutonium, and the subsequent electrodeposition for alpha spectrometry measurement. The technique is based on the use of TBP (tributyl phosphate) as extractant and easy to obtain reactants. It is applicable to various matrices, including water, filters and soils. In the conditions described, is applied to small aliquots of approximately 0.5 g of solid. The technique produces high quality electrodeposits. (authors)

  3. Novel drug delivery systems for actinides (uranium and plutonium) decontamination agents.

    Science.gov (United States)

    Fattal, Elias; Tsapis, Nicolas; Phan, Guillaume

    2015-08-01

    The possibility of accidents in the nuclear industry or of nuclear terrorist attacks makes the development of new decontamination strategies crucial. Among radionuclides, actinides such as uranium and plutonium and their different isotopes are considered as the most dangerous contaminants, plutonium displaying mostly a radiological toxicity whereas uranium exhibits mainly a chemical toxicity. Contamination occurs through ingestion, skin or lung exposure with subsequent absorption and distribution of the radionuclides to different tissues where they induce damaging effects. Different chelating agents have been synthesized but their efficacy is limited by their low tissue specificity and high toxicity. For these reasons, several groups have developed smart delivery systems to increase the local concentration of the chelating agent or to improve its biodistribution. The aim of this review is to highlight these strategies. PMID:26144994

  4. Treatment of Uranium and Plutonium solutions generated in Atalante by R and D activities

    Energy Technology Data Exchange (ETDEWEB)

    Lagrave, H.; Beretti, C.; Bros, P. [CEA Rhone Valley Research Center, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France)

    2008-07-01

    The Atalante complex operated by the 'Commissariat a l'Energie Atomique' (Cea) consolidates research programs on actinide chemistry, processing for recycling spent fuel, and fabrication of actinide targets for innovative concepts in future nuclear systems. In order to produce mixed oxide powder containing uranium, plutonium and minor actinides and to deal with increasing flows in the facility, a new shielded line will be built and is expected to be operational by 2012. Its main functions will be to receive, concentrate and store solutions, purify them, ensure co-conversion of actinides and conversion of excess uranium. (authors)

  5. Reactivity change measurements on plutonium-uranium fuel elements in hector experimental techniques and results

    International Nuclear Information System (INIS)

    The techniques used in making reactivity change measurements on HECTOR are described and discussed. Pile period measurements were used in the majority of oases, though the pile oscillator technique was used occasionally. These two methods are compared. Flux determinations were made in the vicinity of the fuel element samples using manganese foils, and the techniques used are described and an error assessment made. Results of both reactivity change and flux measurements on 1.2 in. diameter uranium and plutonium-uranium alloy fuel elements are presented, these measurements being carried out in a variety of graphite moderated lattices at temperatures up to 450 deg. C. (author)

  6. Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

    International Nuclear Information System (INIS)

    A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs

  7. A review of the corrosion and pyrophoricity behavior of uranium and plutonium

    International Nuclear Information System (INIS)

    This report presents a review of the corrosion and pyrophoricity behavior of uranium and plutonium. For each element, the reactions with oxygen, water vapor, and aqueous solutions are described in terms of reaction rates, products, and mechanisms. Their pyrophoric tendencies in terms of measured ignition temperatures are discussed, and the effects of the important variables specific area, gas composition, and prior storage rare stated. The implications of the observed behavior for current storage issues are considered

  8. Electrolytic Partitioning of Uranium and Plutonium Based on a New Type of Electrolytic Mixer-Settler

    Institute of Scientific and Technical Information of China (English)

    YUAN; Zhong-wei; YAN; Tai-hong; ZHENG; Wei-fang; ZUO; Chen; LI; Hui-rong

    2012-01-01

    <正>A new type of electroreduction mixer-settler for the partitioning of uranium and plutonium during the Purex process, which is featured with E-shaped cathodes and U-shaped anodes in settling chamber, is designed and the operational results achieved using this equipment are presented. The results show that this new type of mixer-settler has excellent separation performance. The flow rate of organic feed solution

  9. Idaho Chemical Processing Plant and Plutonium-Uranium Extraction Plant phaseout/deactivation study

    International Nuclear Information System (INIS)

    The decision to cease all US Department of Energy (DOE) reprocessing of nuclear fuels was made on April 28, 1992. This study provides insight into and a comparison of the management, technical, compliance, and safety strategies for deactivating the Idaho Chemical Processing Plant (ICPP) at Westinghouse Idaho Nuclear Company (WINCO) and the Westinghouse Hanford Company (WHC) Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this study is to ensure that lessons-learned and future plans are coordinated between the two facilities

  10. An intake prior for the Bayesian analysis of plutonium and uranium exposures in an epidemiology study.

    Science.gov (United States)

    Puncher, M; Birchall, A; Bull, R K

    2014-12-01

    In Bayesian inference, the initial knowledge regarding the value of a parameter, before additional data are considered, is represented as a prior probability distribution. This paper describes the derivation of a prior distribution of intake that was used for the Bayesian analysis of plutonium and uranium worker doses in a recent epidemiology study. The chosen distribution is log-normal with a geometric standard deviation of 6 and a median value that is derived for each worker based on the duration of the work history and the number of reported acute intakes. The median value is a function of the work history and a constant related to activity in air concentration, M, which is derived separately for uranium and plutonium. The value of M is based primarily on measurements of plutonium and uranium in air derived from historical personal air sampler (PAS) data. However, there is significant uncertainty on the value of M that results from paucity of PAS data and from extrapolating these measurements to actual intakes. This paper compares posterior and prior distributions of intake and investigates the sensitivity of the Bayesian analyses to the assumed value of M. It is found that varying M by a factor of 10 results in a much smaller factor of 2 variation in mean intake and lung dose for both plutonium and uranium. It is concluded that if a log-normal distribution is considered to adequately represent worker intakes, then the Bayesian posterior distribution of dose is relatively insensitive to the value assumed of M. PMID:24191121

  11. Metallography of plutonium, uranium and thorium fuels: two decades of experience in Radiometallurgy Division

    International Nuclear Information System (INIS)

    Ever since the inception of Radiometallurgy Laboratory (RML) in its early seventies optical metallography has played a key role in development and fabrication of plutonium, uranium and thorium bearing nuclear fuels. In this report, an album of photomicrographs depicts the different types of metallic, ceramic and dispersion fuels and welded section that have been evaluated in RML during the last two decades. (author). 14 refs., 1 tab

  12. Immobilization of uranium and plutonium into boro-basalt, pyroxene and andradite mineral-like compositions

    International Nuclear Information System (INIS)

    The immobilization of plutonium-containing wastes with the manufacturing of stable solid compositions is one of the problems that should be solved in the disposal of radioactive wastes. The works on the choice, preparation with the use of the cold crucible induction melter (CCIM) technology, and investigation of materials that are most suitable for immobilizing plutonium-containing wastes of different origin have been carried out at the All-Russian Scientific Research Institute of Inorganic Materials (VNIINM) and the Institute of the Geology of Ore Deposits, Petrography, Mineralogy, and Geochemistry (IGEM), Russian Academy of Sciences in the framework of the agreements with Lawrence Livermore National Laboratory (LLNL, USA) on the material and technical support. This paper presents the data on the synthesis of cerium-, uranium-, and plutonium-containing materials based on boro-basalt, pyroxene, and andradite compositions in the muffle furnace and by using the CCIM method. The compositions containing up to 15 - 18 wt % cerium oxide, 8 - 11 wt % uranium oxide, and 4.6 - 5.7 wt % plutonium oxide were obtained in laboratory facilities installed in glove boxes. Comparison studies of the materials synthesized in the muffle furnace and CCIM demonstrate the advantages of using the CCIM method. The distribution of components in the materials synthesized are investigated, and their certain physicochemical properties are determined. (authors)

  13. Irradiated uranium reprocessing, Final report I-VI, IV Deo IV - Separation of uranium, plutonium and fission products from the irradiated fuel of the reactor in Vinca

    International Nuclear Information System (INIS)

    This study describes the technology for separation of uranium, plutonium and fission products from the radioactive water solution which is obtained by dissolving the spent uranium fuel from the reactor in Vinca. The procedure should be completed in a hot cell, with the maximum permitted activity of 10 Ci

  14. Investigation of solubility of cesium, strontium, barium, rare-earth, uranium and americium fluorides in acid nitrosyl fluoride (NOFx3HF)

    International Nuclear Information System (INIS)

    Solubility of Am and other elements, which are fission products, in acid nitrosylfluoride has been studied. Cesium fluoride has maximum solubility; uranium tetrafluoride is also noticeably soluble; americium trifluoride is practically insoluble; fluorides of rare earth elements are slightly soluble in NOFx3HF. Analysis of the solid phase obtained after treating the mixture of the above fluorides with acid nitrosylfluoride has shown that cesium fluoride reacts with NOFx3HF with the formation of an acid salt (CsFxHF), whereas fluorides of alkaline and rare earth elements remain unchanged. The behaviour of a mixture of cesium, barium, and lanthanum fluorides in the process of three-multiple treating with acid nitrosylfluoride has been studied. It is shown that more than 98% of cesium fluoride and 5% of barium fluoride pass into the mother liquor while lanthanum fluoride remains completely in the solid phase. The data on americium fluoride solubility in acid nitrosylfluoride have indicated that it behaves in the same way as fluorides of rare earth elements; it is practically insoluble in HOFx3HF

  15. Process for plutonium rextraction in aqueous solution from an organic solvent, especially for uranium plutonium partition

    International Nuclear Information System (INIS)

    The organic solvent containing plutonium is contacted with an aqueous solution of a uranous salt, for instance uranous nitrate, and a hydroxylamine salt, for instance the nitrate. In these conditions uranous nitrate is a reducing agent of Pu III and hydroxylamine nitrate stabilizes Pu III and U IV in the aqueous phase. Performances are similar to these of the U IV-hydrazine nitrate without interference of hydrazine nitrate degradation products

  16. Inhalation toxicology of industrial plutonium and uranium oxide aerosols I. Physical chemical characterization

    International Nuclear Information System (INIS)

    In the fabrication of mixed plutonium and uranium oxide fuel, large quantities of dry powders are processed, causing dusty conditions in glove box enclosures. Inadvertent loss of glove box integrity or failure of air filter systems can lead to human inhalation exposure. Powdered samples and aerosol samples of these materials obtained during two fuel fabrication process steps have been obtained. A regimen of physical chemical tests of properties of these materials has been employed to identify physical chemical properties which may influence their biological behavior and dosimetry. Materials to be discussed are 750 deg. C heat-treated, mixed uranium and plutonium oxides obtained from the ball milling operation and 1750 deg. C heat-treated, mixed uranium and plutonium oxides obtained from the centerless grinding of fuel pellets. Results of x-ray diffraction studies have shown that the powder generated by the centerless grinding of fuel pellets is best described as a solid solution of UOx and PuOx consistent with its temperature history. In vitro dissolution studies of both mixed oxide materials indicate a generally similar dissolution rate for both materials. In one solvent, the material with the higher temperature history dissolves more rapidly. The x-ray diffraction and in vitro dissolution results as well as preliminary results of x-ray photoelectron spectroscopic analyses will be compared and the implications for the associated biological studies will be discussed. (author)

  17. Chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium nitrate solutions

    International Nuclear Information System (INIS)

    These analytical procedures are designed to show whether a given material meets the purchaser's specifications as to plutonium content, effective fissile content, and impurity content. The following procedures are described in detail: plutonium by controlled-potential coulometry; plutonium by amperometric titration with iron(II); free acid by titration in an oxalate solution; free acid by iodate precipitation-potentiometric titration method; uranium by Arsenazo I spectrophotometric method; thorium by thorin spectrophotometric method; iron by 1,10-phenanthroline spectrophotometric method; chloride by thiocyanate spectrophotometric method; fluoride by distillation-spectrophotometric method; sulfate by barium sulfate turbidimetric method; isotopic composition by mass spectrometry; americium-241 by extraction and gamma counting; americium-241 by gamma counting; gamma-emitting fission products, uranium, and thorium by gamma-ray spectroscopy; rare earths by copper spark spectrochemical method; tungsten, niobium (columbium), and tantalum by spectrochemical method; simple preparation by spectrographic analysis for general impurities

  18. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    International Nuclear Information System (INIS)

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

  19. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  20. Crystal chemistry of uranium (V) and plutonium (IV) in a titanate ceramic for disposition of surplus fissile material

    Science.gov (United States)

    Fortner, J. A.; Kropf, A. J.; Finch, R. J.; Bakel, A. J.; Hash, M. C.; Chamberlain, D. B.

    2002-07-01

    We report X-ray absorption near-edge structure (XANES) and extended X-ray absorption fine-structure (EXAFS) spectra for the plutonium LIII and uranium LIII edges in titanate pyrochlore ceramic. The titanate ceramics studied are of the type proposed to serve as a matrix for the immobilization of surplus fissile materials. The samples studied contain approximately 10 wt% fissile plutonium and 20 wt% natural uranium, and are representative of material within the planned production envelope. Based upon natural analogue models, it had been previously assumed that both uranium and plutonium would occupy the calcium site in the pyrochlore crystal structure. While the XANES and EXAFS signals from the plutonium LIII are consistent with this substitution into the calcium site within pyrochlore, the uranium XANES is characteristic of pentavalent uranium. Furthermore, the EXAFS signal from the uranium has a distinct oxygen coordination shell at 2.07 Å and a total oxygen coordination of about 6, which is inconsistent with the calcium site. These combined EXAFS and XANES results provide the first evidence of substantial pentavalent uranium in an octahedral site in pyrochlore. This may also explain the copious nucleation of rutile (TiO 2) precipitates commonly observed in these materials as uranium displaces titanium from the octahedral sites.

  1. Separation by sequential chromatography of americium, plutonium and neptunium elements: application to the study of trans-uranian elements migration in a European lacustrine system

    International Nuclear Information System (INIS)

    The nuclear tests carried out in the atmosphere in the Sixties, the accidents and in particular that to the power station of Chernobyl in 1986, were at the origin of the dispersion of a significant quantity of transuranic elements and fission products. The study of a lake system, such that of the Blelham Tarn in Great Britain, presented in this memory, can bring interesting answers to the problems of management of the environment. The determination of the radionuclides in sediment cores made it possible not only to establish the history of the depositions and consequently the origin of the radionuclides, but also to evaluate the various transfers which took place according to the parameters of the site and the properties of the elements. The studied transuranic elements are plutonium 238, 239-240, americium 241 and neptunium 237. Alpha emitting radionuclides, their determination requires complex radiochemical separations. A method was worked out to successively separate the three radioelements by using a same chromatographic column. Cesium 137 is the studied fission product, its determination is done by direct Gamma spectrometry. Lead 210, natural radionuclide, whose atmospheric flow can be supposed constant. makes it possible to obtain a chronology of the various events. The detailed vertical study of sediment cores showed that the accumulation mode of the studied elements is the same one and that the methods of dating converge. The cesium, more mobile than transuranic elements in the atmosphere, was detected in the 1963 and 1986 fallout whereas an activity out of transuranic elements appears only for the 1963 fallout. The activity of the 1963 cesium fallout is of the same order of magnitude as that of 1986. The calculation of the diffusion coefficients of the elements in the sediments shows an increased migration of cesium compared to transuranic elements. An inventory on the whole of the lake made it possible to note that the atmospheric fallout constitute the

  2. Thermal radiative and thermodynamic properties of solid and liquid uranium and plutonium carbides in the visible-near infrared range

    CERN Document Server

    Fisenko, Anatoliy I

    2016-01-01

    The knowledge of thermal radiative and thermodynamic properties of uranium and plutonium carbides under extreme conditions is essential for designing a new metallic fuel materials for next generation of a nuclear reactor. The present work is devoted to the study of the thermal radiative and thermodynamic properties of liquid and solid uranium and plutonium carbides at their melting/freezing temperatures. The Stefan-Boltzmann law, total energy density, number density of photons, Helmholtz free energy density, internal energy density, enthalpy density, entropy density, heat capacity at constant volume, pressure, and normal total emissivity are calculated using experimental data for the frequency dependence of the normal spectral emissivity of liquid and solid uranium and plutonium carbides in the visible-near infrared range. It is shown that the thermal radiative and thermodynamic functions of uranium carbide have a slight difference during liquid-to-solid transition. Unlike UC, such a difference between these ...

  3. Characterization of past and present solid waste streams from the Plutonium-Uranium Extraction Plant

    International Nuclear Information System (INIS)

    During the next two decades the transuranic wastes, now stored in the burial trenches and storage facilities at the Hanford Site, are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Over 7% of the transuranic waste to be retrieved for shipment to the Waste Isolation Pilot Plant has been generated at the Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this report is to characterize the radioactive solid wastes generated by PUREX using process knowledge, existing records, and oral history interviews. The PUREX Plant is currently operated by the Westinghouse Hanford Company for the US Department of Energy and is now in standby status while being prepared for permanent shutdown. The PUREX Plant is a collection of facilities that has been used primarily to separate plutonium for nuclear weapons from spent fuel that had been irradiated in the Hanford Site's defense reactors. Originally designed to reprocess aluminum-clad uranium fuel, the plant was modified to reprocess zirconium alloy clad fuel elements from the Hanford Site's N Reactor. PUREX has provided plutonium for research reactor development, safety programs, and defense. In addition, the PUREX was used to recover slightly enriched uranium for recycling into fuel for use in reactors that generate electricity and plutonium. Section 2.0 provides further details of the PUREX's physical plant and its operations. The PUREX Plant functions that generate solid waste are as follows: processing operations, laboratory analyses and supporting activities. The types and estimated quantities of waste resulting from these activities are discussed in detail

  4. Characterization of past and present solid waste streams from the Plutonium-Uranium Extraction Plant

    Energy Technology Data Exchange (ETDEWEB)

    Pottmeyer, J.A.; Weyns, M.I.; Lorenzo, D.S.; Vejvoda, E.J. [Los Alamos Technical Associates, Inc., NM (US); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (US)

    1993-04-01

    During the next two decades the transuranic wastes, now stored in the burial trenches and storage facilities at the Hanford Site, are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Over 7% of the transuranic waste to be retrieved for shipment to the Waste Isolation Pilot Plant has been generated at the Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this report is to characterize the radioactive solid wastes generated by PUREX using process knowledge, existing records, and oral history interviews. The PUREX Plant is currently operated by the Westinghouse Hanford Company for the US Department of Energy and is now in standby status while being prepared for permanent shutdown. The PUREX Plant is a collection of facilities that has been used primarily to separate plutonium for nuclear weapons from spent fuel that had been irradiated in the Hanford Site`s defense reactors. Originally designed to reprocess aluminum-clad uranium fuel, the plant was modified to reprocess zirconium alloy clad fuel elements from the Hanford Site`s N Reactor. PUREX has provided plutonium for research reactor development, safety programs, and defense. In addition, the PUREX was used to recover slightly enriched uranium for recycling into fuel for use in reactors that generate electricity and plutonium. Section 2.0 provides further details of the PUREX`s physical plant and its operations. The PUREX Plant functions that generate solid waste are as follows: processing operations, laboratory analyses and supporting activities. The types and estimated quantities of waste resulting from these activities are discussed in detail.

  5. Impact of MCNP unresolved resonance probability-table treatment on uranium and plutonium benchmarks

    International Nuclear Information System (INIS)

    Versions of MCNP up through and including MCNP 4B have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into a developmental version of MCNP. MCNP results are presented for a variety of uranium and plutonium critical benchmarks, calculated with and without the probability-table treatment. The results from these calculations, along with their associated standard deviations, are presented. Four conclusions can be drawn from these results. First, not surprisingly, the only benchmarks that are substantially affected are those that have a significant fraction of their interactions within the unresolved resonance region of the principal uranium and plutonium isotopes that are present. For example, the unreflected metal spheres and the graphite-reflected IEU sphere have spectra that are too hard to produce large reactivity changes, while the fluoride and nitrate solutions have spectra that are too thermal. Second, the reactivity impact of the improvement is essentially negligible for 235U in these systems. The spectral peaks for all of the HEU benchmarks except GODIVA occur within the unresolved resonance region, but the probability-table treatment does not produce a statistically significant change in reactivity for any of them. Third, the probability-table method can produce substantial increases in reactivity for those benchmarks that include proportionately large amounts of 238U and high fluxes within the unresolved resonance region. Finally, the probability-table method also can produce significant reactivity changes for plutonium benchmarks with intermediate spectra. However, the magnitude of those changes is considerably smaller than for some of the cases with 238U. There are competing reactivity effects among the different plutonium isotopes, and in addition, the unresolved resonance range for the plutonium isotopes is not as broad as that for 238U

  6. Calibration of X-ray densitometers for the determination of uranium and plutonium concentrations in reprocessing input and product solutions

    International Nuclear Information System (INIS)

    In June 1985 a calibration exercise has been carried out, which included the calibration of the KfK K-Edge Densitometer for uranium assay in the uranium product solutions from reprocessing, and the calibration of the Hybrid K-Edge/K-XRF Instrument for the determination of total uranium and plutonium in reprocessing input solutions. The calibration measuremnts performed with the two X-ray densitometers are described and analyzed, and calibration constants are evaluated from the obtained results. (orig.)

  7. Uranium, plutonium... Radionuclides - Which effects on the living?

    International Nuclear Information System (INIS)

    After a recall of the definitions of some important words and notions related to radioactivity and irradiation, a first article discusses the development of biokinetic and dosimetric models for the description of radioactive elements in the human body, and the energy deposition due to their disintegration. Main French and international actors are indicated. A second article discusses the role of bones which may fix an exogenous metal and thus act as a defence of the body by preventing this metal to damage more sensitive soft organs. A third article addresses the evolution of the study of uranium toxicology, and more particularly discusses the biochemical study of the mechanisms of interaction between uranyl (the solute form of uranium) and proteins: the objective is to better understand its transport, its bio-distribution, its retention, its mode of action and its excretion. The fourth article addresses the elaboration of molecules which could trap radionuclides and actinides within the body and limit their toxicity in case of contamination due to accidental exposure. The fifth article addresses the search for tools which could be used to characterize molecular physical-chemical interactions of radio-elements. The next article addresses chelation as a principle for contamination remediation in the case of intoxication by actinides: the author gives an overview of chemical and biological constraints in the design of these new chelation treatments

  8. Gastrointestinal absorption of plutonium and uranium in fed and fasted adult baboons and mice: application to humans

    International Nuclear Information System (INIS)

    Gastrointestinal (GI) absorption values of plutonium and uranium were determined in fed and fasted adult baboons and mice. For both baboons and mice, the GI absorptions of plutonium and uranium were 10 to 20 times higher in 24 h fasted animals than in fed ones. For plutonium, GI absorption values in baboons were almost identical to those in mice for both fed and fasted conditions, and values for fed animals agreed with estimates for humans. For uranium, GI absorption values in fed and fasted baboons were 6 to 7 times higher than those in mice, and agreed well with those fed and fasted humans. For one baboon that was not given its morning meal, plutonium absorption 2 h after the start of the active phase was the same as that in the 24 h fasted animals. In contrast, for baboons that received a morning meal, plutonium absorption did not rise to the value of 24 h fasted baboons even 8 h after the meal. We conclude that GI absorption values for plutonium and uranium in adult baboons are good estimates of the values in humans and that the values for the fasted condition should be used to set standards for oral exposure of persons in the workplace. (author)

  9. Evaluation of organic and inorganic adsorbents for the removal of uranium and plutonium from process streams

    Energy Technology Data Exchange (ETDEWEB)

    Herald, W.R.; Koenst, J.W.; Luthy, D.F.

    1977-01-01

    Mound Laboratory is evaluating macroporous, ion exchange resins for the removal of plutonium, uranium, and various colloids from process waste treatment effluents. A number of organic ion exchange resins were evaluated for removal of /sup 238/Pu(IV), /sup 238/Pu(VI), and /sup 233/U(VI) from water using batch isotherm tests. The capacity and equilibrium distribution coefficients were compared with each other and with bone char, an inorganic adsorbent consisting of hydroxyapatite (HAP). The various types of adsorbents showed that the extent of removal and the equilibrium coefficients (Kd) were functions of pH. For removal of polymeric plutonium, /sup 238/Pu(IV), the best results were achieved using the inorganic adsorbent, bone char (hydroxyapatite), at pH 7. However, macroporous, weak base, anion exchange resins also showed reasonable Kd values at pH 7. Therefore, the best removal of polymeric plutonium can be achieved using chemisorption or weak base anionic exchange, indicating strongly ionized anions. Excellent results for removal of /sup 238/Pu(VI) were achieved using macroporous, strong base, anion exchange resins and macroporous, strong acid, cation exchange resins. For removal of ionic /sup 233/U(VI), the strongly acidic cation exchangers gave the better results; the Kd values were on the order of 10/sup 2/ better than bone char. Again, performance was strongly dependent upon pH. Adsorbent resins which remove constituents by physical adsorption did not perform well for uranium removal.

  10. Studies of plutonium-iron and uranium-plutonium-iron alloys

    International Nuclear Information System (INIS)

    We study the plutonium-iron system, by means of dilatometry, X rays and metallography, especially in the domain between PuFe2 and Fe. We determine the solubilities of Fe in PuFe2 and of Pu in Fe. We show the presence of an hexagonal PuFe2 phase and we propose a modification in the Pu-Fe phase diagram. Some low iron concentration U-Pu-Fe alloys have also been investigated. We characterise the different phases. We confirm that adding some iron lowers the quantity of the zeta U-Pu phase. We emphasize some characteristics of the alloys having the global concentration (U, Pu)6 Fe. (authors)

  11. Concentration of uranium and plutonium in unsaturated spent fuel tests

    International Nuclear Information System (INIS)

    Commercial spent fuel is being tested under oxidizing conditions at 90 C in drip tests with simulated groundwater to evaluate its long-term performance in a potential repository at Yucca Mountain [1-4]. The tests allow us to monitor the dissolution behavior of the spent fuel matrix and the release rates of individual radionuclides. This paper reports the U and Pu concentrations in the leachates of drip tests during 3.7 years of reaction. Changes in these concentrations are correlated with changes in the measured pH and the appearance of alteration products on the fuel surface. Although there is little thermodynamic information at 90 C for either uranyl or plutonium compounds, some data are available at 25 C [5-8]. The literature data for the U and Pu solubilities of U and Pu compounds were compared to the U and Pu concentrations in the leachates. We also compare Wilson's [9] U and Pu concentrations in semi-static tests at 85 C on spent fuel with our results

  12. Status of Americium-241 recovery at Rocky Flats Plant

    International Nuclear Information System (INIS)

    This paper is presented in two parts: Part I, Molten Salt Extraction of Americium from Molten Plutonium Metal, and Part II, Aqueous Recovery of Americium from Extraction Salts. The Rocky Flats recovery process used for waste salts includes (1) dilute hydrochloric acid dissolution of residues; (2) cation exchange to convert from the chloride to the nitrate system and to remove gross amounts of monovalent impurities; (3) anion exchange separation of plutonium; (4) oxalate precipitation of americium; and (5) calcination of the oxalate at 6000C to yield americium oxide. The aqueous process portion describes attempts to improve the recovery of americium. The first part deals with modifications to the cation exchange step; the second describes development of a solvent extractions process that will recovery americium from residues containing aluminium as well as other common impurities. Results of laboratory work are described. 3 figures, 6 tables. (DP)

  13. Plutonium

    International Nuclear Information System (INIS)

    This report contains with regard to 'plutonium' statements on chemistry, occurrence and reactions in the environment, handling procedures in the nuclear fuel cycle, radiation protection methods, biokinetics, toxicology and medical treatment to make available reliable data for the public discussion on plutonium especially its use in nuclear power plants and its radiological assessment. (orig.)

  14. Reaction of uranium and plutonium carbides with nitrogen; Reaction avec l'azote des carbures d'uranium et de plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzelli, R.; Martin, A.; Schickel, R. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1966-03-01

    Uranium and plutonium carbides react with nitrogen during the grinding process preceding the final sintering. The reaction occurs even in argon atmospheres containing a few percent of residual nitrogen. The resulting contamination is responsible for the appearance of an equivalent quantity of higher carbide in the sintered products; nitrogen remains quantitatively in the monocarbide phase. UC can be transformed completely into nitride under a nitrogen pressure, at a temperature as low as 400 C. The reaction is more sluggish with PuC. The following reactions take places: UC + 0,8 N{sub 2} {yields}> UN{sub 1.60} + C and PuC + 0,5 N{sub 2} {yields} PuN + C. (authors) [French] Les carbures d'uranium et de plutonium reagissent avec l'azote au cours du broyage qui precede le frittage final. Cette reaction est sensible meme sous des atmospheres d'argon ne contenant que quelques pour cent d'azote. Cette contamination se traduit sur les produits frittes par l'apparition d'une quantite equivalente de carbure superieur, l'azote restant fixe quantitativement dans la phase monocarbure. On peut transformer entierement UC en nitrure par action de l'azote sous pression des 400 C. La reaction est plus difficile avec PuC. Les reactions sont les suivantes: UC + 0,8 N{sub 2} {yields} UN{sub 1.60} + C et PuC + 0,5 N{sub 2} {yields} PuN + C.

  15. An improvement in APOR process I-uranium/plutonium separation process

    Institute of Scientific and Technical Information of China (English)

    肖松涛; 李丽; 叶国安; 罗方祥; 刘协春; 杨贺; 兰天

    2015-01-01

    The reduction stripping behavior of Pu(IV) from 30%TBP/OK with hydroxysemicarbazide (HSC) was inves-tigated, and the separation efficiency of HSC and DMHAN-MMH for U/Pu partitioning in Purex process was compared. The results show that HSC can effectively realize the separation of Pu from U;using mixer-settlers to simulate U/Pu separation in 1B bank of PUREX, from 16-stage counter current extraction experiment (in which 6 stages for supplemental extraction, 10 stages for stripping) with flow rate ratio (1BF:1BX:1BS)=4:1:1 in 1B contactor, good result was achieved that the yields are both more than 99.99%for uranium and Pu, the separation factor of plutonium from uranium (SFPu/U) is 2.8 × 104, and separation factor of uranium from plu-tonium (SFU/Pu) is 5.9 × 104. As a stripping reductant, HSC can effectively achieve the separation of Pu from U and the separation effect is nearly the same with DMHAN-MMH, which contributed to replace enough the latter with HSC in the U/Pu separation in Advanced Purex Process Based on Organic Reagent (APOR) process.

  16. Plutonium and uranium emission experience in U.S. nuclear facilities using HEPA filtration

    International Nuclear Information System (INIS)

    The weekly, monthly and yearly emission experience of eleven U.S. nuclear facilities is presented, and reviewed. Each of these facilities uses HEPA filtration to control emissions, the various vent streams using from one to three such filtration systems in series. Available emission records cover from one to five vents at each facility, and cover periods ranging from two to eleven years. The majority of the facilities process plutonium, either exclusively or as mixed oxide, but a few process highly enriched uranium-235, natural uranium and/or thorium. Facilities include fuel fabrication plants, fuel reprocessing plants, and experimental facilities. The emission curves show clearly not only the normal operation of the filter systems, but also the effects of minor leaks, and of major failures, as well as the effects of retention of material in ducts and other passive parts of the filtration systems. An analysis is presented of the operation of such systems in real environments, and the non-Stokes Law behavior of very dense particles is exemplified by experiments done with a non-radoactive tracer, whose crystal is isomorphous with those of plutonium and uranium dioxides, tungsten dioxide

  17. Calculation of critical parameters for uranium and/or plutonium nitrate solution, 2

    International Nuclear Information System (INIS)

    For the purpose of criticality experiments of the nitrate solution systems composed of uranium and/or plutonium, the designs of Criticality Safety Experimental Facility(CSEF) are under-going in JAERI. In this report, by using the developed calculational code for the atomic number density of the nitrate solution of U/Pu mixture, the ratio H/Fissile has been calculated as a function of fuel concentration and acidity. Critical parameters such as infinite multiplication factor, effective multiplication factor, and the critical diameter of infinite cylinder have been evaluated with Monte Carlo code KENO-IV of JACS system. The dependence of multiplication factors on the formula for the atomic number density is described by comparing the calculated results with our formula and those with ARH-600 formula. Based on our formula, the effects of acid molality and Plutonium valence state on the criticality of nitrate solution have been discussed. (author)

  18. Plutonium-uranium separation in the Purex process using mixtures of hydroxylamine nitrate and ferrous sulfamate

    International Nuclear Information System (INIS)

    Laboratory studies, followed by plant operation, established that a mixture of hydroxylamine nitrate (HAN) and ferrous sulfamate (FS) is superior to FS used alone as a reductant for plutonium in the Purex first cycle. FS usage has been reduced by about 70% (from 0.12 to 0.04M) compared to the pre-1978 period. This reduced the volume of neutralized waste due to FS by 194 liters/metric ton of uranium (MTU) processed. The new flowsheet also gives lower plutonium losses to waste and at least comparable fission product decontamination. To achieve satisfactory performance at this low concentration of FS, the acidity in the 1B mixer-settler was reduced by using a split-scrub - a low acid scrub in stage one and a higher acid scrub in stage three - to remove acid from the solvent exiting the 1A centrifugal contactor. 8 references, 14 figures, 1 table

  19. Separation of trace amount of uranium from the dissolved solution of plutonium oxide

    International Nuclear Information System (INIS)

    A solvent extraction method has been developed for the separation of trace concentration of uranium (3-30μg/ml) from plutonium solution (10-12mg/ml) obtained on dissolution of plutonium oxide in HNO3 containing traces of HF. The method involves the extraction of U(VI) with mixed solvent comprising 2% trialkyl phosphine oxide (TRPO) - 1.1M tri-n-butyl phosphate (TBP) - 0.4 M tertiary butyl hydroquinone (TBH) solution in n-dodecane from 3MHNO3 - 0.6M N2H4, while reducing Pu to inextractable Pu(III) in the aqueous phase. Decontamination factor of ∼ 85 against Pu and recovery better than 92% for U (at 5-15μg) have been achieved. (author)

  20. Method to manufacture nuclear fuel monocarbide, particularly uranium and uranium-plutonium monocarbide

    International Nuclear Information System (INIS)

    Uranium monocarbide and U-Pu monocarbides are manufactured by converting the oxides with uranium carbide (mol ratio 3:1) at temperatures >16000C in vacuum (-6 bar). Only small quantities of CO are formed, one obtains high-density pellets which are very suitable as nuclear fuels. (IHOE)

  1. Ultraslow Wave Nuclear Burning of Uranium-Plutonium Fissile Medium on Epithermal Neutrons

    CERN Document Server

    Rusov, V D; Eingorn, M V; Chernezhenko, S A; Kakaev, A A

    2014-01-01

    For a fissile medium, originally consisting of uranium-238, the investigation of fulfillment of the wave burning criterion in a wide range of neutron energies is conducted for the first time, and a possibility of wave nuclear burning not only in the region of fast neutrons, but also for cold, epithermal and resonance ones is discovered for the first time. For the first time the results of the investigation of the Feoktistov criterion fulfillment for a fissile medium, originally consisting of uranium-238 dioxide with enrichments 4.38%, 2.00%, 1.00%, 0.71% and 0.50% with respect to uranium-235, in the region of neutron energies 0.015-10.0eV are presented. These results indicate a possibility of ultraslow wave neutron-nuclear burning mode realization in the uranium-plutonium media, originally (before the wave initiation by external neutron source) having enrichments with respect to uranium-235, corresponding to the subcritical state, in the regions of cold, thermal, epithermal and resonance neutrons. In order to...

  2. Toxicity of uranium and plutonium to the developing embryos of fish. [Cyprinus carpio, Pimephales promelas

    Energy Technology Data Exchange (ETDEWEB)

    Till, J.E.; Kaye, S.V.; Trabalka, J.R.

    1976-07-01

    The radiological and chemical toxicity of plutonium and uranium to the developing embryos of fish was investigated using eggs from carp, Cyprinus carpio, and fathead minnows, Pimephales promelas. Freshly fertilized eggs were developed in solutions containing high specific activity /sup 238/Pu or /sup 232/U or low specific activity /sup 244/Pu, /sup 235/U, or /sup 238/U. Quantitative tests to determine the penetration of these elements through the chorion indicated that plutonium accumulated in the contents of carp eggs reaching a maximum concentration factor of approximately 3.0 at hatching. Autoradiographs of 16 ..mu.. egg sections showed that plutonium was uniformly distributed in the egg volume. Uranium localized in the yolk material, and the concentration factor in the yolk sac remained constant during development at approximately 3.3. Doses from /sup 238/Pu which affected hatchability of the eggs were estimated to be 1.6 x 10/sup 4/ rads and 9.7 x 10/sup 3/ rads for C. carpio and P. promelas, respectively; doses from /sup 232/U were 1.3 x 10/sup 4/ rads for C. carpio and 2.7 x 10/sup 3/ rads for P. promelas. A greater number of abnormal larvae than in control groups was produced by /sup 238/Pu doses of 4.3 x 10/sup 3/ rads to carp and 5.7 x 10/sup 2/ rads to fathead minnows; 3.2 x 10/sup 3/ rads and 2.7 x 10/sup 2/ rads were estimated from /sup 232/U. Eggs that were incubated in 20 ppM /sup 244/Pu did not hatch. This mortality may have been the result of chemical toxicity of plutonium. Concentrations of 60 ppM of /sup 235/U and /sup 238/U did not affect egg hatching. Based on these data, concentrations in fish eggs were calculated for representative concentrations of uranium and plutonium in natural waters and the corresponding dose levels are below those levels at which observable effects begin to occur.

  3. Design, construction and performance test of the analysis lines of mixed uranium-plutonium carbide and nitride fuels for LMFBR

    International Nuclear Information System (INIS)

    Design, construction and performance test of the analysis lines of uranium, plutonium, carbon, nitrogen and oxygen in mixed uranium-plutonium carbide and nitride fuels are described. Since the fuels are chemically unstable in the air, they are sealed before analysis in metallic capsules using dies and oil pressure in a high-purity argon atmosphere glove box. The capsules are then transferred respective apparatuses installed in four air-atmosphere glove boxes and the elements are determined. Commercial apparatuses were drastically modified for safe plutonium handling and easy maintenance in the glove boxes. Performance test with uranium compounds showed that the elements of the fuels could be determined successfully by the analysis lines. The methods developed of determining oxygen and nitrogen in oxide and nitride fuels containing them in high percentages are also described. (author)

  4. Calculation of criticality parameters for uranium and/or plutonium nitrate solutions, 1

    International Nuclear Information System (INIS)

    For the purpose of computing criticality parameters of the solutions with uranium and plutonium, the atomic number density was formulated and programmed in the computer. The amount of solvent in the solution was calculated from solute concentration and density of the solution. The numerical expression of the solution density was based on the literature data for aqueous nitrate solution, and on theoretical consideration for 30% TBP-n. dodecane solution. The calculated values of solution density were discussed compared with that in ARH-600, the criticality handbook in the United States. (author)

  5. Method of reprocessing of irradiated nuclear fission products of the uranium, plutonium and thorium group

    International Nuclear Information System (INIS)

    A solvent extraction is used to separate irradiated nuclear fission materials of the group uranium, plutonium, thorium from radioactive fission products which are present together in an aqueous solution. An improvement on the known mehod is proposed in which a carboxylic nitrile, carboxylic ester, carboxylic amide, or a mixture of these substances is added to the organic phase which is mixed with a non-polar diluting agent as a polar modificator, where the modificators are derived from mono- or polycarboxylic acids or also from substituted carboxylic acids. Amyl acetate, N-N dimethyl caprylic acid amide, and adiponitrile are particularly suitable. (UW/LH)

  6. Measurement of prompt neutron decay constant on uranium-plutonium metal assembly

    International Nuclear Information System (INIS)

    The prompt neutron decay constant α of the Uranium-Plutonium metal assembly was measured with Rossi-α technique at delay critical and sub-critical conditions, and then αc is obtained. The relationship between α with reactivity and α is analyzed with detector counts reciprocal. The comparison shows that the direct measured result under delay critical condition is 0.835 ± 0.005/μs, which error is 1/6 of that from Rossi-a and extrapolation methods. (authors)

  7. Gastrointestinal absorption of plutonium, uranium and neptunium in fed and fasted adult baboons: Application to humans

    International Nuclear Information System (INIS)

    Gastrointestinal (GI) absorption values of plutonium, uranium, and neptunium were determined in fed and fasted adult baboons. A dual isotope method of determining GI absorption, which does not require animal sacrifice, was validated and shown to compare well with the sacrifice method (summation of oral isotope in urine with that in tissues at sacrifice). For all three elements, mean GI absorption values were significantly high (5- to 50-fold) in 24-hour (h)-fasted animals than in fed animals, and GI absorption values for baboons agreed well with those for humans

  8. Standard guide for determination of plutonium and neptunium in uranium hexafluoride by alpha spectrometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This method covers the determination of plutonium and neptunium isotopes in uranium hexafluoride by alpha spectroscopy. The method can also be applicable to any matrix that may be converted to a nitric acid system. 1.2 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory requirements prior to use.

  9. Incentives for transmutation of americium in thermal reactors

    International Nuclear Information System (INIS)

    This report describes possible benefits when americium is irradiated in a thermal reactor. If all plutonium is partitioned from spent fuel, americium is the main contributor to the radiotoxicity of spent fuel upto several thousands of years of storage. It is shown that americium can be transmuted to other nuclides upon irradiation in a thermal reactor, leading to a 50% reduction of the radiotoxicity of neptunium, which can be an important contributor to the dose due to leakage of nuclides after one million years of storage. The radiotoxicity of americium can be reduced considerably after irradiation for 3 to 6 years in a thermal reactor with thermal neutron flux of 1014 cm-2s-1. The strongly α and neutron emitting transmutation products can most probably not be recycled again, so a transmutation process is suggested in which americium is irradiated for 3 to 6 years and then put to final storage. It is shown that the radiotoxicity of the transmuation products after a storage time of about one hundred years can be considerably reduced compared to the radiotoxicity of the initial americium. The same holds for the α activity and heat emission of the transmutation products. Because plutonium in spent fuel contributes for about 80% to the radiotoxicity upto 105 years of storage, recycling and transmutation of plutonium has first priority. Transmutation of americium is only meaningful when the radiotoxicity of plutonium is reduced far below the radiotoxicity of americium. (orig.)

  10. Fallout of uranium and plutonium from recent volcanic eruptions

    International Nuclear Information System (INIS)

    The concentrations of 234U, 235U, 238U in rain water were measured in a total of 102 individual samples which were collected at Fayetteville, Arkansas, from July 1980 through April 1983. A spectacular increase in the heavy isotope of uranium (238U) was observed in the months of July, August (1980); January through April, November and December 1981. This large increase in 238U in rain appeared to have had its origin in the May 1980 eruption of Mount St. Helens. An increase in the concentration of 238U in rain, smaller than 1981, was observed, which seems to have originated from the El Chichon volcano eruption in March 1982, and the spring peak or so-called cycling effect. A striking increase in the average bimonthly concentration of /sup 239,240/Pu occurred during the months of September-October 1980 (15.6 fCi/l) and March-April 1981 (29.4 fCi/l). The excess deposition of /sup 239,240/Pu brought down by the rain at Fayetteville, Arkansas, from March 1980 through December 1982 was found to be 1.01 fCi/cm2. The total amount of /sub 239,240/Pu deposited at Fayetteville, Arkansas, from March through December 1982, was found to be about 30 times higher than the total amount calculated from reported literature values. The excess /sup 239,240/Pu has been attributed to stratospheric /sub 239,240/Pu from nuclear weapons testing prior to the 25th Chinese nuclear test

  11. Thermal conductivity of uranium-plutonium oxide fuel for fast reactors

    International Nuclear Information System (INIS)

    A new thermal conductivity correlation for fully dense uranium-plutonium oxide fuel for fast reactors was formulated for fuel pin thermal analysis under beginning of irradiation conditions. The data set used in correlating the equation was systematically selected to minimize experimental uncertainty. The electron conduction term for uranium dioxide formulated by Harding and Martin [J. Nucl. Mater. 166 (1989) 223] was adopted to compensate for so few high temperature measurements. The excellent predictability of the new correlation was validated by comparing the calculated with measured fuel center temperatures in an instrumented irradiation test in the experimental fast reactor JOYO for low oxygen-to-metal (O/M) ratio fuel up to 1850 K

  12. Plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Miner, William N

    1964-01-01

    This pamphlet discusses plutonium from discovery to its production, separation, properties, fabrication, handling, and uses, including use as a reactor fuel and use in isotope power generators and neutron sources.

  13. Uranium and plutonium total half-lives and for the spontaneous fission branch

    Energy Technology Data Exchange (ETDEWEB)

    Holden, N.E.

    1985-01-01

    The long-lived nuclides of the uranium and plutonium elements are of interest for their use in nuclear reactors, as well as in certain safeguard applications, e.g., alpha counting is often used to determine the amount of material present. The total half-life and the half-life for spontaneous fission are evaluated for these various long-lived nuclides of interest. The various experiments have been reanalyzed and recommended values are presented for /sup 232,233,234/U, /sup 235,236,238/U, and for /sup 236,238,239,240,241,242,244/Pu. These values improve upon preliminary estimates previously presented, in particular with respect to the uncertainties reported. The /sup 234/U half-life of 2.456 +- 0.005 x 10/sup 5/ years impacts directly on the 2200 meters/second fission cross section of /sup 235/U, since earlier measurements used values of 2.47 to 2.5 x 10/sup 5/ years and obtained correspondingly lower cross sections. In a similar manner, the /sup 239/Pu half-life is 1.25% lower than earlier estimates, which results in a 1.25% increase in the 2200 m/s fission cross section for /sup 239/Pu in some earlier cross section measurements. The total half-lives for the uranium nuclides were reviewed some time ago. At that time, the only spontaneous fission value which was evaluated was /sup 238/U. Recently, the uranium and plutonium nuclides were reviewed for both total and fission half-lives. The general procedure followed in this paper has been to review each of the experiments and revise the published values for the latest estimates of the various parameters used by the original authors. For the case of the total half-lives of uranium, only differences from the original work have been discussed. 120 refs., 23 tabs.

  14. Summary of active test of uranium-plutonium co-denitration facility at Rokkasho reprocessing plant

    International Nuclear Information System (INIS)

    The aim of this report is to explain and discuss the active test results in the uranium-plutonium (U-Pu) co-denitration facility. We had previously performed the uranium test with depleted uranium from February of 2005 to January of 2006. Then, the active test has been in progress since March of 2006 toward the start of commercial operation. Plutonium nitrate (PuN) and uranium nitrate hexahydrate (UNH) are mixed at the ratio of approximately 1:1 from the non-proliferation viewpoint. The mixed solution is supplied into the denitration dish inside the denitration oven where the solution is denitrated by microwave heating and converted to MOX powder (PuO2-UO3). After denitration, the powder is converted to the product of MOX powder (PuO2-UO2) through some heating processes and stored in temporary canisters. The powder is transferred to the blender, and then filled into powder cans. 3 powder cans are packed into a canister and transferred to storage in the co-denitrated product powder storage building. Confirmation of the denitration ability of the mixed solution and characteristics of the product powder, (1) Stable and continuous operation in the target period, (2) Characteristics of the product powder, (3) Processing ability at each process, (4) Impurities in the product powder. The test results of the last step of the active test of the U-Pu co-denitration facility are presented; (1) Average throughput in 5 days at A and B lines was more than the target value. (2) Mean particle sizes and specific surface areas in MOX powder were within the standards. (3) Each process indicated good result. (4) Impurities in product powder were less than each limitation. (author)

  15. Uranium and plutonium total half-lives and for the spontaneous fission branch

    International Nuclear Information System (INIS)

    The long-lived nuclides of the uranium and plutonium elements are of interest for their use in nuclear reactors, as well as in certain safeguard applications, e.g., alpha counting is often used to determine the amount of material present. The total half-life and the half-life for spontaneous fission are evaluated for these various long-lived nuclides of interest. The various experiments have been reanalyzed and recommended values are presented for /sup 232,233,234/U, /sup 235,236,238/U, and for /sup 236,238,239,240,241,242,244/Pu. These values improve upon preliminary estimates previously presented, in particular with respect to the uncertainties reported. The 234U half-life of 2.456 +- 0.005 x 105 years impacts directly on the 2200 meters/second fission cross section of 235U, since earlier measurements used values of 2.47 to 2.5 x 105 years and obtained correspondingly lower cross sections. In a similar manner, the 239Pu half-life is 1.25% lower than earlier estimates, which results in a 1.25% increase in the 2200 m/s fission cross section for 239Pu in some earlier cross section measurements. The total half-lives for the uranium nuclides were reviewed some time ago. At that time, the only spontaneous fission value which was evaluated was 238U. Recently, the uranium and plutonium nuclides were reviewed for both total and fission half-lives. The general procedure followed in this paper has been to review each of the experiments and revise the published values for the latest estimates of the various parameters used by the original authors. For the case of the total half-lives of uranium, only differences from the original work have been discussed. 120 refs., 23 tabs

  16. Strong association of fallout plutonium with humic and fulvic acid as compared to uranium and 137Cs in Nishiyama soils from Nagasaki, Japan

    International Nuclear Information System (INIS)

    To investigate the formation of mobile organic plutonium, the plutonium contents of the fulvic (FA) and humic (HA) acids were analyzed from the soil samples obtained at Nishiyama, Nagasaki, Japan. The percentages of the plutonium bound strongly to HA and to FA vs. the total plutonium in the soil were 5-10% and 1%, respectively, at the depth of 0-0.1 m, much higher values than those of 137Cs and uranium. After being weathered for 51 years under a temperate climate, the initial highfired oxides of fallout plutonium have become as chemically reactive plutonium from nuclear fuel reprocessing plants. (author)

  17. Extraction of hexavalent uranium, tetravalent plutonium and fission products by N, N'-tetraalkyldiamides

    International Nuclear Information System (INIS)

    This study deals with the extractive properties of N, N'-tetraalkylglutaramides of generic formula R2NC(0)(CH2)3C(0)NR2. These molecules were considered as alternative extractants to tributylphosphate in nuclear fuels reprocessing. They are selective extractants of uranium and plutonium as far as trivalent actinides and lanthanides remain in aqueous nitric solutions. Distribution ratios measurements and F.T. Infra-Red investigations show that HN03 extraction takes place via the formation of the following species: 2L.HN03, L.HN03 and L.2HN03 in the organic phase (L: glutaramide). Distribution ratios of actinide ions followed by UV-visible spectroscopy and Infra-Red investigations agree with formation of the following neutral organometallic complexes in low nitric acidity conditions: L.U02(N03)2 and L.Pu(N03)4 and the anionic species at higher acidities: L.U02(N03)3H and L.Pu(N03)6H2. Interactions occur through neutral complexes and free molecules of diamides which explain the non ideality of the organic phase. Degradation products of these molecules don't seem to alter the extractive properties of these extractants towards uranium and plutonium

  18. Dose Contribution from High Level Waste Uranium and Plutonium. Revision 1

    International Nuclear Information System (INIS)

    Radiological source terms for safety analyses traditionally have been curie lists of radionuclides. Converting the source term to dose values allows each radionuclide to be evaluated for its impact on dose, which is the purpose of the source term. This report is one in a series of reports establishing source terms for High Level Waste (HLW) by evaluating the dose impact of each radionuclide. These reports will be used in establishing the source terms to be used in HLW Safety Analysis Reports. The purpose of this report is to document the bounding element dose impact of uranium and plutonium in HLW. This technique (use of dose rather than curies) demonstrates vividly the relative importance of these nuclides in accident analyses. A large amount of available data permitted dose values to be established for uranium and plutonium; therefore, these two elements were evaluated independent of other nuclides. Solubility and adsorption data, available for these elements, allow bounding conditions to be established for their contribution to dose for various HLW processes

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

    Directory of Open Access Journals (Sweden)

    Golubev V.K.

    2012-08-01

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

  20. Pulverizing and mixing method and device for uranium/plutonium mixed oxide powder

    International Nuclear Information System (INIS)

    The present invention concerns a uranium/plutonium mixed oxide fuel (MOX) and it provides a device for uniformly mixing PuO2 powder and UO2 powder. Namely, a vessel contains a uranium/plutonium mixed oxide powder at a thickness of lower than a critical level. An upper disk and a lower disk are opposed to each other at a predetermined distance in the inside of the vessel. Each of the disks has crusher teeth in an annular shape with a different diameter. An oxide powder is supplied to the vessel, and the upper disk and the lower disk are rotated at a high speed in the direction opposite to each other. As is described above, since the oxide powder is mixed under pulverization while being dispersed by the method and the device of the present invention, the powder is mixed at a good homogeneity. Since the oxide powder is pulverized and mixed in the vessel, the operation can be preceded in a completely sealed state. Since the operation can be conducted continuously, the processing performance can be improved. Since the thickness of the powder in the inside of the vessel and the thickness thereof at a discharge port are at a subcritical level, control for the critical safety is facilitated. (I.S.)

  1. Separation Of Uranium And Plutonium Isotopes For Measurement By Multi Collector Inductively Coupled Plasma Mass Spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Martinelli, R E; Hamilton, T F; Williams, R W; Kehl, S R

    2009-03-29

    Uranium (U) and plutonium (Pu) isotopes in coral soils, contaminated by nuclear weapons testing in the northern Marshall Islands, were isolated by ion-exchange chromatography and analyzed by mass spectrometry. The soil samples were spiked with {sup 233}U and {sup 242}Pu tracers, dissolved in minerals acids, and U and Pu isotopes isolated and purified on commercially available ion-exchange columns. The ion-exchange technique employed a TEVA{reg_sign} column coupled to a UTEVA{reg_sign} column. U and Pu isotope fractions were then further isolated using separate elution schemes, and the purified fractions containing U and Pu isotopes analyzed sequentially using multi-collector inductively coupled plasma mass spectrometer (MCICP-MS). High precision measurements of {sup 234}U/{sup 235}U, {sup 238}U/{sup 235}U, {sup 236}U/{sup 235}U, and {sup 240}Pu/{sup 239}Pu in soil samples were attained using the described methodology and instrumentation, and provide a basis for conducting more detailed assessments of the behavior and transfer of uranium and plutonium in the environment.

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

    Science.gov (United States)

    Golubev, V. K.

    2012-08-01

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

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

    Science.gov (United States)

    Golubev, Vladimir

    2015-06-01

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

  4. Isotopic composition and origin of uranium and plutonium in selected soil samples collected in Kosovo

    International Nuclear Information System (INIS)

    Soil samples collected from locations in Kosovo where depleted uranium (DU) ammunition was expended during the 1999 Balkan conflict were analysed for uranium and plutonium isotopes content (234U, 235U, 236U, 238U, 238Pu, 239+240Pu). The analyses were conducted using gamma spectrometry (235U, 238U), alpha spectrometry (238Pu, 239+240Pu), inductively coupled plasma-mass spectrometry (ICP-MS) (234U, 235U, 236U, 238U) and accelerator mass spectrometry (AMS) (236U). The results indicated that whenever the U concentration exceeded the normal environmental values (∼2 to 3 mg/kg) the increase was due to DU contamination. 236U was also present in the released DU at a constant ratio of 236U (mg/kg)/238U (mg/kg)=2.6x10-5, indicating that the DU used in the ammunition was from a batch that had been irradiated and then reprocessed. The plutonium concentration in the soil (undisturbed) was about 1 Bq/kg and, on the basis of the measured 238Pu/239+240Pu, could be entirely attributed to the fallout of the nuclear weapon tests of the 1960s (no appreciable contribution from DU)

  5. Isotopic composition and origin of uranium and plutonium in selected soil samples collected in Kosovo

    Energy Technology Data Exchange (ETDEWEB)

    Danesi, P.R. E-mail: P.R.Danesi@iaea.org; Bleise, A.; Burkart, W.; Cabianca, T.; Campbell, M.J.; Makarewicz, M.; Moreno, J.; Tuniz, C.; Hotchkis, M

    2003-07-01

    Soil samples collected from locations in Kosovo where depleted uranium (DU) ammunition was expended during the 1999 Balkan conflict were analysed for uranium and plutonium isotopes content ({sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U, {sup 238}Pu, {sup 239+240}Pu). The analyses were conducted using gamma spectrometry ({sup 235}U, {sup 238}U), alpha spectrometry ({sup 238}Pu, {sup 239+240}Pu), inductively coupled plasma-mass spectrometry (ICP-MS) ({sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U) and accelerator mass spectrometry (AMS) ({sup 236}U). The results indicated that whenever the U concentration exceeded the normal environmental values ({approx}2 to 3 mg/kg) the increase was due to DU contamination. {sup 236}U was also present in the released DU at a constant ratio of {sup 236}U (mg/kg)/{sup 238}U (mg/kg)=2.6x10{sup -5}, indicating that the DU used in the ammunition was from a batch that had been irradiated and then reprocessed. The plutonium concentration in the soil (undisturbed) was about 1 Bq/kg and, on the basis of the measured {sup 238}Pu/{sup 239+240}Pu, could be entirely attributed to the fallout of the nuclear weapon tests of the 1960s (no appreciable contribution from DU)

  6. Plutonium-uranium mixed oxide characterization by coupling micro-X-ray diffraction and absorption investigations

    Science.gov (United States)

    Degueldre, C.; Martin, M.; Kuri, G.; Grolimund, D.; Borca, C.

    2011-09-01

    Plutonium-uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The potential differences of metal redox state and microstructural developments of the matrix before and after irradiation are commonly analysed by electron probe microanalysis. In this work the structure and next-neighbor atomic environments of Pu and U oxide features within unirradiated homogeneous MOX and irradiated (60 MW d kg -1) MOX samples was analysed by micro-X-ray fluorescence (μ-XRF), micro-X-ray diffraction (μ-XRD) and micro-X-ray absorption fine structure (μ-XAFS) spectroscopy. The grain properties, chemical bonding, valences and stoichiometry of Pu and U are determined from the experimental data gained for the unirradiated as well as for irradiated fuel material examined in the center of the fuel as well as in its peripheral zone (rim). The formation of sub-grains is observed as well as their development from the center to the rim (polygonization). In the irradiated sample Pu remains tetravalent (>95%) and no (oxidation in the rim zone. Any slight potential plutonium oxidation is buffered by the uranium dioxide matrix while locally fuel cladding interaction could also affect the redox of the fuel.

  7. Criticality Parameters for Mixtures of Plutonium Oxide, Uranium Oxide and Water

    International Nuclear Information System (INIS)

    A number of articles have appeared in the literature in 1964-5, concerned with various aspects of reactor fuels containing mixtures of plutonium and uranium. So far as the author is aware only one of these, a report by Marchuk, is concerned with criticality. The present paper considers homogeneous mixtures of plutonium and uranium oxides with water in various proportions and presents estimates of the critical parameters of these mixtures based upon diffusion theory calculations using four energy groups. No experimental results are available to directly check the accuracy of these calculations. Using the same method of calculation, however, the experimental results quoted for 0,372 in. diam. rods in hexagonal lattice arrangement in water were reproduced to within ±1.1% in keff. These rods contained 1.5 wt.% PuO2 in UO2 depleted to 0.22 wt.%235U. The results of some Monte Carlo check calculations on homogeneous mixtures are quoted ; these predict lower critical masses. Comparison is also made with Marchuk's results. The mixtures considered in this paper are assumed to be mechanical mixtures of water with PuO2/UO2 having a crystal density of 11.2 g/cm3. Marchuk's calculations are based on a PuO2/UO2 density of 6.0 g/cm3. The comparisons with his results assume the same density. (author)

  8. Characterization of uranium and plutonium in surface-waters and sediments collected at the Rocky Flats Facility

    International Nuclear Information System (INIS)

    This study was initiated to characterize actinides in environmental samples collected at the Rocky Flats Plant (RFP). Thermal Ionization Mass Spectrometry (TIMS) measurement techniques were used to measure the plutonium and uranium content of water and sediment samples collected from the ponds used to control surface-waters on-site at RFP. TIMS was also used to separate the uranium into anthropogenic and naturally occurring components. The results of these studies are presented

  9. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Secretariat has received a note verbale dated 20 September 2012 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2011. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of highly enriched uranium (HEU) as of 31 December 2011

  10. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Director General has received a note verbale dated 14 October 2010 from the Permanent Mission of the Federal Republic of Germany to the IAEA in enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2009. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil high enriched uranium (HEU) as of 31 December 2009

  11. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of Highly Enriched Uranium

    International Nuclear Information System (INIS)

    The Director General has received a Note Verbale dated 3 July 2007 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2006. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil highly enriched uranium (HEU) as of 31 December 2006

  12. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Director General has received a note verbale dated 29 April 2011 from the Permanent Mission of the Federal Republic of Germany to the IAEA in enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2010. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil high enriched uranium (HEU) as of 31 December 2010

  13. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Secretariat has received a note verbale dated 2 July 2013 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2012. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of highly enriched uranium (HEU) as of 31 December 2012

  14. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Director General has received a letter dated 16 July 2009 from the Permanent Mission of the Federal Republic of Germany to the IAEA in enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2008. 2. The Government of the Federal Republic of Germany has also made available a statement of its annual figures for holdings of civil high enriched uranium (HEU) as of 31 December 2008

  15. Uranium and/or plutonium nitride preparation proces for nuclear fuels. Procede de preparation de nitrure d'uranium et/ou de plutonium utilisable comme combustible nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Bardelle, P.; Bernard, H.

    1989-03-17

    Uranium and/or plutonium mononitride(s) is (are) obtained by nitruration of corresponding oxide(s) in presence of carbon. Oxygen and carbon contents are reduced by use of controlled atmospheres first nitrogen, then hydrogenated rare gas (hydrogen < 8 vol%) during heating and hydrogenated rare gas during cooling.

  16. Theoretical methods for determination of core parameters in uranium-plutonium lattices

    International Nuclear Information System (INIS)

    The prediction of plutonium production in power reactors depends essentially on how the change of neutron energy spectra in a reactor cell during burn-up is determined. In the epithermal region, where the build-up of plutonium occurs, the slowing down effects are particularly important, whereas, on the other hand, the thermal neutron spectrum is strongly influenced by the low-lying plutonium resonances. For accurate analysis, multi-group numerical methods are required, which, applied to burn-up prediction, are extremely laborious and time consuming even for large computers. This paper contains a comprehensive review of the methods of core parameter determination in the uranium-plutonium lattices developed in Yugoslavia during the last few years. Faced with the problem of using small computers, the authors had to find new approaches combining physical evidence and mathematical elegance. The main feature of these approaches is the tendency to proceed with analytical treatment as far as possible and then to include suitable numerical improvements. With this philosophy, which is generally overlooked when using large computers, fast and reasonably accurate methods were developed. The methods include original means for adequate treatment of neutron spectra and cell geometry effects,especially suitable for U-Pu systems. In particular, procedures based on the energy dependent boundary conditions, the discrete energy representation, the improved collision probabilities and the Green function slowing down solutions were developed and applied. Results obtained with these methods are presented and compared with those of the experiments and those obtained with other methods. (author)

  17. Nonproliferation and safeguards aspects of fuel cycle programs in reduction of excess separated plutonium and high-enriched uranium

    International Nuclear Information System (INIS)

    The purpose of this preliminary investigation is to explore alternatives and strategies aimed at the gradual reduction of the excess inventories of separated plutonium and high-enriched uranium (HEU) in the civilian nuclear power industry. The study attempts to establish a technical and economic basis to assist in the formation of alternative approaches consistent with nonproliferation and safeguards concerns. Reference annual mass flows and inventories for a representative 1,400 Mwe Pressurized Water Reactor (PWR) fuel cycle have been investigated for three cases: the 100 percent uranium oxide UO2 fuel loading once through cycle, and the 33 percent mixed oxide MOX loading configuration for a first and second plutonium recycle. The analysis addresses fuel cycle developments; plutonium and uranium inventory and flow balances; nuclear fuel processing operations; UO2 once-through and MOX first and second recycles; and the economic incentives to draw-down the excess separated plutonium stores. The preliminary analysis explores several options in reducing the excess separated plutonium arisings and HEU, and the consequences of the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials on nonproliferation and safeguards policy assessments

  18. Evaluation of an L/sub III/ x-ray absorption-edge densitometer for assay of mixed uranium-plutonium solutions

    International Nuclear Information System (INIS)

    An L/sub III/ x-ray absorption-edge densitometer (XRAED), designed and built at Los Alamos Scientific Laboratory, has been evaluated at Savannah River Laboratory for the assay of uranium and plutonium in process solutions. Fro 2000-second data collection, the precisions (95% confidence level) for uranium assays varied +- 1.04% at 10 g of uranium per liter to 0.34% at 45 g/L, and plutonium assays varied from +- 6.7% at 2 g of plutonium per liter to +- 2.2% at 10 g/L. Hydroxylamine nitrate, a reducing agent used in solvent extraction tests, affect the accuracies of both uranium and plutonium assay by densitometry, and plutonium analysis by coulometry. It appears that the XRAED could be used to control the SRL miniature mixer-settlers, but additional testing will be needed to demonstrate that XRAED accuracy is sufficient for accountability of special nuclear materials

  19. Methods for analysis of uranium-plutonium mixed fuel and transplutonium elements at RIAR (Preprint no. IT-25)

    International Nuclear Information System (INIS)

    Different methods for analysis of the uranium-plutonium mixed nuclear fuel and transplutonium elements are briefly discussed in this paper: coulometry, radiometric techniques, emission spectrography, mass-spectrometry, chromatography, spectrophotometry. The main analytical characteristics of the methods developed are given. (author). 30 refs., 2 tabs

  20. Safeguards design strategies: designing and constructing new uranium and plutonium processing facilities in the United States

    International Nuclear Information System (INIS)

    In the United States, the Department of Energy (DOE) is transforming its outdated and oversized complex of aging nuclear material facilities into a smaller, safer, and more secure National Security Enterprise (NSE). Environmental concerns, worker health and safety risks, material security, reducing the role of nuclear weapons in our national security strategy while maintaining the capability for an effective nuclear deterrence by the United States, are influencing this transformation. As part of the nation's Uranium Center of Excellence (UCE), the Uranium Processing Facility (UPF) at the Y-12 National Security Complex in Oak Ridge, Tennessee, will advance the U.S.'s capability to meet all concerns when processing uranium and is located adjacent to the Highly Enriched Uranium Materials Facility (HEUMF), designed for consolidated storage of enriched uranium. The HEUMF became operational in March 2010, and the UPF is currently entering its final design phase. The designs of both facilities are for meeting anticipated security challenges for the 21st century. For plutonium research, development, and manufacturing, the Chemistry and Metallurgy Research Replacement (CMRR) building at the Los Alamos National Laboratory (LANL) in Los Alamos, New Mexico is now under construction. The first phase of the CMRR Project is the design and construction of a Radiological Laboratory/Utility/Office Building. The second phase consists of the design and construction of the Nuclear Facility (NF). The National Nuclear Security Administration (NNSA) selected these two sites as part of the national plan to consolidate nuclear materials, provide for nuclear deterrence, and nonproliferation mission requirements. This work examines these two projects independent approaches to design requirements, and objectives for safeguards, security, and safety (3S) systems as well as the subsequent construction of these modern processing facilities. Emphasis is on the use of Safeguards-by-Design (SBD

  1. The United States transuranium and uranium registries (USTUR). Learning from plutonium and uranium workers

    International Nuclear Information System (INIS)

    Beginning in the 1960's with the mission of acquiring and providing precise information about the effects of plutonium and other transuranic elements in man, the USTUR has followed up to 'old age' almost 500 volunteer Registrants who worked at weapons sites and received measurable internal doses. While failing (despite careful life-time follow-up) to demonstrate deleterious health effects attributable to transuranic elements, USTUR research, based on these real human data from DOE workers, continues its contributions to the development of the biokinetic models used internationally to assess intakes from bioassay data and predict tissue doses. There is still much to learn from the Registries '370 deceased tissue donors and the 110 still-living Registrants, whose average age is now about 76 years (youngest 95 y). This paper illustrates USTUR's current 5-y research program, including the application of registrant case data to (i) quantify the variability in behavior of transuranic materials among individuals; (ii) validate new methodologies used at DOE sites for assessing 'realistic' tissue doses in individual cases; and (iii) model the effectiveness of chelation therapy. These data can also be used to examine the adequacy of protection standards utilized for plutonium workers in the early years of the nuclear industry. (author)

  2. LITERATURE REVIEW ON THE SORPTION OF PLUTONIUM, URANIUM, NEPTUNIUM, AMERICIUM AND TECHNETIUM TO CORROSION PRODUCTS ON WASTE TANK LINERS

    Energy Technology Data Exchange (ETDEWEB)

    Li, D.; Kaplan, D.

    2012-02-29

    The Savannah River Site (SRS) has conducted performance assessment (PA) calculations to determine the risk associated with closing liquid waste tanks. The PA estimates the risk associated with a number of scenarios, making various assumptions. Throughout all of these scenarios, it is assumed that the carbon-steel tank liners holding the liquid waste do not sorb the radionuclides. Tank liners have been shown to form corrosion products, such as Fe-oxyhydroxides (Wiersma and Subramanian 2002). Many corrosion products, including Fe-oxyhydroxides, at the high pH values of tank effluent, take on a very strong negative charge. Given that many radionuclides may have net positive charges, either as free ions or complexed species, it is expected that many radionuclides will sorb to corrosion products associated with tank liners. The objective of this report was to conduct a literature review to investigate whether Pu, U, Np, Am and Tc would sorb to corrosion products on tank liners after they were filled with reducing grout (cementitious material containing slag to promote reducing conditions). The approach was to evaluate radionuclides sorption literature with iron oxyhydroxide phases, such as hematite ({alpha}-Fe{sub 2}O{sub 3}), magnetite (Fe{sub 3}O{sub 4}), goethite ({alpha}-FeOOH) and ferrihydrite (Fe{sub 2}O{sub 3} {center_dot} 0.5H{sub 2}O). The primary interest was the sorption behavior under tank closure conditions where the tanks will be filled with reducing cementitious materials. Because there were no laboratory studies conducted using site specific experimental conditions, (e.g., high pH and HLW tank aqueous and solid phase chemical conditions), it was necessary to extend the literature review to lower pH studies and noncementitious conditions. Consequently, this report relied on existing lower pH trends, existing geochemical modeling, and experimental spectroscopic evidence conducted at lower pH levels. The scope did not include evaluating the appropriateness of K{sub d} values for the Fe-oxyhydroxides, but instead to evaluate whether it is a conservative assumption to exclude this sorption process of radionuclides onto tank liner corrosion products in the PA model. This may identify another source for PA conservatism since the modeling did not consider any sorption by the tank liner.

  3. Concentration and speciation of plutonium, americium, uranium, thorium, potassium and 137Cs in a venice canal sediment sample

    Science.gov (United States)

    Testa, C.; Desideri, D.; Guerra, F.; Meli, M. A.; Roselli, C.; Degetto, S.

    1999-01-01

    A sequential extraction method consisting of six operationally-defined fractions has been developed for determining the geochemical partitioning of natural (U, Th, 40K) and antropogenic (Pu, Am, 137Cs) radionuclides in a 40-50 cm deep sediment sample collected in a Venice canal. Extraction chromatography with Microthene-TOPO (U, Th), Microthene-TNOA (Pu) and Microthene-HDEHP (Am) column was used for the chemical separation of a single radionuclide; the final recoveries were calculated by adding 236U, 229Th, 242Pu and 243Am as the yield tracers. After electrodeposition the alpha spectrometry was carried out. 137Cs and 40K were measured by gamma spectrometry. The total concentrations in the wet sample (Bq/kgd), obtained by a complete disgregation of the matrix by wet and dry treatment, were the following: 239+240Pu=1.03±0.07, 238Pu=0.022±0.005, 241Am=0.337±0.027, 137Cs=9.78±0.78, 238U=28.84±1.62, 232Th=21.42±1.93, 40K=376.05±12.78. The mean ratio 238Pu/239+240Pu (0.02) shows a contamination due essentially to fall-out and U and Th alpha spectra indicate the natural origin of two elements. The absence of 134Cs in the sample proves that at 40-50 cm depth the sediment was not affected by the Chernobyl fall-out. As far as the speciation is concerned the following fractions were considered: water soluble, carbonates, Fe-Mn oxides, organic matter, acid soluble, residue. Pu (˜67%) an Am (˜95%) were present principally in the carbonate fraction; U was more distributed and about 30% and 45% appeared in the carbonate fraction and in the residue respectively; the majority of Th was present in the residue (˜60%); 40K was totally present in the residue; finally 137Cs was found mostly in the acid soluble fraction (˜53%) and in the residue (˜47%). Some stable elements (Fe, Mn, Al, Ti, Ca, Pb, Ba) were also determined in the different fractions to get more information about the chemical association of the single radionuclides.

  4. Burnup behavior of FBR fuels sourced in uranium and plutonium recycled in PWRs and its influence on fuel cycle economy

    International Nuclear Information System (INIS)

    Considering the strategic security of uranium resources, authors investigated the effective use of fuel recycling in the current light water reactor system (1.1GWe-class PWRs of standard type), where we supposed the uranium fuel reformed by re-enrichment of the recovery uranium from the PWR spent fuels and the MOX fuel produced with the recovery plutonium as main fissile, and examined three types of MOX fuels prepared by making choice of these matrixes among (A) the recovery uranium, (B) the depleted uranium as the waste from natural uranium enrichment, and (C) the depleted one sourced from recovery uranium re-enrichment. Calculations were carried out of the multi-component uranium isotope separation employing the ideal centrifuge-cascade and of the reactor core burnup analysis using the comprehensive neutronics computation system SRAC-2006. Burnup analysis was performed under the following conditions: The concentration of 235U in the depleted natural uranium is 0.2 mol%. This conditional mass-effect on isotope separation affects the other isotope concentrations in the centrifuge cascade. The integrated burnup of uranium and full-MOX fuels is 45 GWd/t- HM during 3 cycles in one batch burning. The spent original-uranium-fuels are cooled for 10 years before these reprocessing and then fabrication of reformed uranium or MOX fuels. In the reprocessing, the plutonium is recovered as a 50-50 mixture with the spent uranium oxide and the remaining uranium oxide is recovered in isolated form. The result of burnup analysis shows that the uranium recovered from the spent original fuel contains 235U enriched about 20 % more than that of the natural uranium and also 236U transformed from 235U capturing neutrons during fuel burning. The constituent 236U behaves as a neutron absorber. Hence, the reformed uranium fuel containing 236U enriched additionally in the re-enrichment process requires the fissile 235U concentrated 1.154 times more than that in the original fuel, in order

  5. Crystal growth and first crystallographic characterization of mixed uranium(IV)-plutonium(III) oxalates.

    Science.gov (United States)

    Tamain, Christelle; Arab Chapelet, Bénédicte; Rivenet, Murielle; Abraham, Francis; Caraballo, Richard; Grandjean, Stéphane

    2013-05-01

    The mixed-actinide uranium(IV)-plutonium(III) oxalate single crystals (NH4)0.5[Pu(III)0.5U(IV)0.5(C2O4)2·H2O]·nH2O (1) and (NH4)2.7Pu(III)0.7U(IV)1.3(C2O4)5·nH2O (2) have been prepared by the diffusion of different ions through membranes separating compartments of a triple cell. UV-vis, Raman, and thermal ionization mass spectrometry analyses demonstrate the presence of both uranium and plutonium metal cations with conservation of the initial oxidation state, U(IV) and Pu(III), and the formation of mixed-valence, mixed-actinide oxalate compounds. The structure of 1 and an average structure of 2 were determined by single-crystal X-ray diffraction and were solved by direct methods and Fourier difference techniques. Compounds 1 and 2 are the first mixed uranium(IV)-plutonium(III) compounds to be structurally characterized by single-crystal X-ray diffraction. The structure of 1, space group P4/n, a = 8.8558(3) Å, b = 7.8963(2) Å, Z = 2, consists of layers formed by four-membered rings of the two actinide metals occupying the same crystallographic site connected through oxalate ions. The actinide atoms are nine-coordinated by oxygen atoms from four bidentate oxalate ligands and one water molecule, which alternates up and down the layer. The single-charged cations and nonbonded water molecules are disordered in the same crystallographic site. For compound 2, an average structure has been determined in space group P6/mmm with a = 11.158(2) Å and c = 6.400(1) Å. The honeycomb-like framework [Pu(III)0.7U(IV)1.3(C2O4)5](2.7-) results from a three-dimensional arrangement of mixed (U0.65Pu0.35)O10 polyhedra connected by five bis-bidentate μ(2)-oxalate ions in a trigonal-bipyramidal configuration. PMID:23577593

  6. Optimized Chemical Separation and Measurement by TE TIMS Using Carburized Filaments for Uranium Isotope Ratio Measurements Applied to Plutonium Chronometry.

    Science.gov (United States)

    Sturm, Monika; Richter, Stephan; Aregbe, Yetunde; Wellum, Roger; Prohaska, Thomas

    2016-06-21

    An optimized method is described for U/Pu separation and subsequent measurement of the amount contents of uranium isotopes by total evaporation (TE) TIMS with a double filament setup combined with filament carburization for age determination of plutonium samples. The use of carburized filaments improved the signal behavior for total evaporation TIMS measurements of uranium. Elevated uranium ion formation by passive heating during rhenium signal optimization at the start of the total evaporation measurement procedure was found to be a result from byproducts of the separation procedure deposited on the filament. This was avoided using carburized filaments. Hence, loss of sample before the actual TE data acquisition was prevented, and automated measurement sequences could be accomplished. Furthermore, separation of residual plutonium in the separated uranium fraction was achieved directly on the filament by use of the carburized filaments. Although the analytical approach was originally tailored to achieve reliable results only for the (238)Pu/(234)U, (239)Pu/(235)U, and (240)Pu/(236)U chronometers, the optimization of the procedure additionally allowed the use of the (242)Pu/(238)U isotope amount ratio as a highly sensitive indicator for residual uranium present in the sample, which is not of radiogenic origin. The sample preparation method described in this article has been successfully applied for the age determination of CRM NBS 947 and other sulfate and oxide plutonium samples. PMID:27240571

  7. Recycling of nuclear matters. Myths and realities. Calculation of recycling rate of the plutonium and uranium produced by the French channel of spent fuel reprocessing

    International Nuclear Information System (INIS)

    The recycling rate of plutonium and uranium are: from the whole of the plutonium separated from the spent fuel ( inferior to 1% of the nuclear matter content) attributed to France is under 50% (under 42 tons on 84 tons); from the whole of plutonium produced in the French reactors is less than 20% (42 tons on 224 tons); from the whole of the uranium separated from spent fuels attributed to France is about 10 % (1600 tons on 16000 tons); from the whole of the uranium contained in the spent fuel is slightly over 5%. (N.C.)

  8. Energy dispersive X ray fluorescence with graphite monochromator - uranium and plutonium analyses in aqueous or organic media

    International Nuclear Information System (INIS)

    We describe an energy dispersive X ray fluorescence apparatus equipped with a cylinder graphite monochromator developed in our laboratory. The graphite monochromator is inserted between the sample and the detector, it permits the selection of the fluorescence X radiation from the sample before collection by the Si-Li diode. Hence, the signal versus noise ratio of the fluorescence peak is increased and the limit of detection of our apparatus for uranium is 0.1 mg/l. This apparatus is perfectly adapted for the L rays of all the transuranium elements determinations. We also demonstrate the possibility of determination of uranium or plutonium in either aqueous or organic (TBP) phase. This apparatus is well adapted to control low levels of uranium and plutonium solutions originating from Purex process

  9. Degrading the Plutonium Produced in Fast Breeder Reactor Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jor-Shan; Kuno, Yusuke [Tokyo University, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan)

    2009-06-15

    Plutonium quality, defined as the plutonium isotopic composition, is an important measure for proliferation-resistance (PR) of a nuclear energy system. The quality of the plutonium produced in the blanket assemblies of a fast breeder reactor could be as good as or better than the weapons-grade (WG). The presence of such good quality plutonium is a proliferation concern. There are various options to degrade the plutonium produced in the breeder blanket. The obvious one is to blend the blanket plutonium with those produced from the reactor core during reprocessing. Other options try to prevent the generation of good quality plutonium (Pu). The Protected Plutonium Production (P{sup 3}) Project proposed by Tokyo Institute of Technology (TIT)1,2,3 advocates the doping of certain amount of neptunium (Np), or americium (Am) in fresh blanket fuel for irradiation. The increased production of {sup 238}Pu, {sup 240}Pu and {sup 242}Pu by neutron capture in {sup 237}Np and Am would degrade the blanket plutonium. However, as {sup 237}Np is a controlled material according to IAEA, its use as doping material in fresh blanket fuel presents a concern for nuclear proliferation. In addition, the fabrication of fresh blanket fuel with inclusion of americium would be complicated due to the emission of intense low-energy gamma radiation from {sup 241}Am. Am is normally accompanied by Cm since the separation of those 2 elements is very difficult. Fuel containing both Am and Cm may make Safeguards measurement difficult. A variation would be doping the fresh blanket fuel with minor actinide (e.g., a group of neptunium, americium, and curium), or with separated reactor-grade (RG) plutonium. The drawback of such schemes would be the need for glove boxes in fresh blanket fuel fabrication. It is possible to fuel the breeder blankets with recycled (reprocessed) uranium oxide. The recycled uranium, recovered from reprocessing, contains {sup 236}U, which when irradiated in the blanket would

  10. Study on the method and instrument for rapid determination of uranium, plutonium and neptunium in spent fuel reprocessing solution by flow injection analysis

    International Nuclear Information System (INIS)

    Based on the principle of flow injection analysis (FIA), a special instrument for measuring uranium, plutonium and neptunium is developed; three rapid analytical methods for measuring uranium, plutonium and neptunium are established respectively. The sensitivity of these methods is -1 μg/ml, the precision is better than 4%. It takes about 25 min to measure the concentration of plutonium or neptunium one time, and 40 s to measure the concentration of uranium at a time. The instrument is composed of the work box connected with radioactive solution and the control box connected with non-radioactive solution to facilitate safety; carrier solution and color reagent are pumped into by a synchronous single plunger pump, thus eliminating pulsation of the baseline; two alternative flow routes are used for measuring plutonium (or neptunium) and uranium respectively. A low pressure chromatographic column is installed in the flow route of determining plutonium or neptunium to monitor enrichment of plutonium or neptunium and to separate other interference elements. In the flow route of determining uranium, a three-way mixer is installed for adding buffer solution and masking agent. Arizona's (III) acts as color reagent for determining uranium, plutonium, and neptunium. The described instrument and method have been used in the analysis of practical samples

  11. Post irradiation examinations of 87F-2A capsule containing uranium-plutonium mixed carbide fuels

    International Nuclear Information System (INIS)

    One fuel pin filled with hyperstoichiometric uranium-plutonium mixed carbide pellets was encapsulated in 87F-2A and irradiated in JMTR up to 4.4 %FIMA at an average linear power of 60 kW/m. The capsule cooled for ∼4 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. It was found from the radial cross section of fuel pin that the helium gap between the pellets and the cladding tube was completely closed. Compared with the results obtained so far, very low open porosity and fission gas release rate as well as mild restructuring was observed owing to the adoption of thermally stable pellets. The diametric increase of fuel pin reached ∼0.06mm at the position of maximum reading, although it might not affect the fuel performance itself. The inner surface of cladding tube did not show signs of carburization. (author)

  12. Three-dimensional microstructural characterization of bulk plutonium and uranium metals using focused ion beam technique

    Science.gov (United States)

    Chung, Brandon W.; Erler, Robert G.; Teslich, Nick E.

    2016-05-01

    Nuclear forensics requires accurate quantification of discriminating microstructural characteristics of the bulk nuclear material to identify its process history and provenance. Conventional metallographic preparation techniques for bulk plutonium (Pu) and uranium (U) metals are limited to providing information in two-dimension (2D) and do not allow for obtaining depth profile of the material. In this contribution, use of dual-beam focused ion-beam/scanning electron microscopy (FIB-SEM) to investigate the internal microstructure of bulk Pu and U metals is demonstrated. Our results demonstrate that the dual-beam methodology optimally elucidate microstructural features without preparation artifacts, and the three-dimensional (3D) characterization of inner microstructures can reveal salient microstructural features that cannot be observed from conventional metallographic techniques. Examples are shown to demonstrate the benefit of FIB-SEM in improving microstructural characterization of microscopic inclusions, particularly with respect to nuclear forensics.

  13. Chemical interaction in uranium-plutonium mixed oxide fuel pins for LMFBR

    International Nuclear Information System (INIS)

    A review is made on the current understanding and problems of chemical interaction between uranium-plutonium mixed oxide and stainless steel cladding for LMFBR fuel pins. The oxygen potential of the fuel was considered as one of the key factors that influences the interaction and the methods of its measurement, its change with irradiation, effect of oxygen redistribution and measured values of irradiated fuel are described. The mechanisms of conventional intergranular and matrix attacks and more recent cladding component chemical transport (CCCT), which was proposed by GE and has been often observed in highly irradiated fuel pins, are explained. Finally, description is given on a statistical analysis of the attack depth and method of inhibiting the cladding. (author)

  14. Safety of Uranium and Plutonium Mixed Oxide Fuel Fabrication Facilities. Specific Safety Guide

    International Nuclear Information System (INIS)

    This Safety Guide supplements the Safety Requirements publication Safety of Fuel Cycle Facilities and addresses all the stages in the life cycle of MOX fuel fabrication facilities, with emphasis placed on design and operation. It describes the actions, conditions and procedures for meeting safety requirements and deals specifically with the handling, processing and storage of plutonium oxide, depleted, natural or reprocessed uranium oxide or mixed oxide manufactured from the above to be used as a feed material to form MOX fuel rods and assemblies for export and subsequent use in water reactors and fast breeder reactors. The publication is intended to be of use to designers, operating organizations and regulators to ensure the safety of MOX fuel fabrication facilities. Contents: 1. Introduction; 2. General safety recommendations; 3. Site evaluation; 4. Design; 5. Construction; 6. Commissioning; 7. Operation; 8. Decommissioning; Annexes.

  15. Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

  16. Sorption of cesium, radium, protactinium, uranium, neptunium and plutonium on rapakivi granite

    Energy Technology Data Exchange (ETDEWEB)

    Huitti, T.; Hakanen, M. [Helsinki Univ. (Finland). Lab. of Radiochemistry; Lindberg, A. [Geological Survey of Finland, Espoo (Finland)

    1996-12-01

    The aim of the study is to determine the sorption of cesium, radium, protactinium, uranium, neptunium and plutonium on rapakivi granite in the brackish groundwater of Haestholmen (site of the Loviisa-1, Loviisa-2 reactors). The studies were carried out under aerobic (Cs, Ra, Pa, U, Np, Pu) and anaerobic (Np, Pa, Pu, Tc) laboratory conditions. The cation exchange capasity was determined for the rock and the diffusion of tritiated water in the rocks of different degree of alteration. The sorption and diffusion properties of the rocks are briefly compared with those of host rocks at other sites under investigation by the Finnish company Posiva Oy for the final disposal of spent fuel. (29 refs.).

  17. Thermodynamic functions and vapor pressures of uranium and plutonium oxides at high temperatures

    International Nuclear Information System (INIS)

    The total energy release in a hypothetical reactor accident is sensitive to the total vapor pressure of the fuel. Thermodynamic functions which are accurate at high temperature can be calculated with the methods of statistical mechanics provided that needed spectroscopic data are available. This method of obtaining high-temperature vapor pressures should be greatly superior to the extrapolation of experimental vapor pressure measurements beyond the temperature range studied. Spectroscopic data needed for these calculations are obtained from infrared spectroscopy of matrix-isolated uranium and plutonium oxides. These data allow the assignments of the observed spectra to specific molecular species as well as the calculation of anharmonicities for monoxides, bond angles for dioxides, and molecular geometries for trioxides. These data are then employed, in combination with data on rotational and electronic molecular energy levels, to determine thermodynamic functions that are suitable for the calculation of high-temperature vapor pressures

  18. Tables of thermodynamic functions for gaseous thorium, uranium, and plutonium oxides

    International Nuclear Information System (INIS)

    Measured and estimated spectroscopic data for thorium, uranium, and plutonium oxide vapor species have been used with the methods of statistical mechanics to calculate thermodynamic functions. Some inconsistencies between spectroscopic data and some thermodynamic data have been resolved by recalculating ΔH0/sub f/ (298.150K) values for the vapor species of these oxides. Evaluation of the uncertainties in data, methods of estimating molecular parameters, and effects of assumptions have been discussed elsewhere. The tables of thermodynamic functions that were reported earlier have been revised principally because the low-frequency vibrational modes of UO2 and UO3 have now been measured. These new empirical data resulted in changes in the electronic contributions to the calculated thermodynamic functions of UO2 and the estimated vibrational contributions for PuO2. In addition, some minor changes have been made in the methods of calculation of the electronic contributions for all molecules

  19. Fission gas release of uranium-plutonium mixed nitride and carbide fuels

    International Nuclear Information System (INIS)

    Uranium-plutonium mixed nitride and carbide for advanced fast reactor fuels were irradiated at JRR-2, and the fission gas release from these fuels were determined. It is confirmed from the irradiation tests that the application of the cold fuel concept to these fuels on the basis of their advantageous thermal properties may realize low fission gas release. Furthermore, the introduction of the thermal stable pellets with dense matrix and relatively large pores can lower the fission gas release to a few percent up to the burnup of 5.5% FIMA. In spite of the retention of fission gas release in the thermal stable pellets, no significant enhancement of the fuel-clad mechanical interaction was observed in the examined range of burnup. It is also suggested that the open porosity would strongly influence the fission gas release from nitride and carbide. (author). 18 refs, 6 figs, 6 tabs

  20. The uranium and plutonium Doppler-effect investigation in the fast critical assemblies

    International Nuclear Information System (INIS)

    The results of experimental investigations into the reactivity Doppler effects in 238 U and Pu by means of measuring the reactivity coefficients for cold (27 deg C) and heated (600 deg C) samples of depleted uranium dioxide and plutonium realized in several BFS and KBR critical assemblies are considered. The method of computational analysis ensuring the comparison between the measuring results and calculational data obtained using algorithms taking into account the factors of resonance self-shielding for the BNAB system of group constants and their Doppler increments is described. Basing on the results of comparison the measuring results for 238 U with calculational data the conclusion is made that the approach discussed gives an opportunity to take into account the Doppler increments in the forbidden resonance range

  1. Investigation on uranium and plutonium nitrides with low oxygen and carbon contents

    International Nuclear Information System (INIS)

    Uranium nitride (UN) and uranium-plutonium nitride (UO.8Pu0.2N) with various oxygen impurity levels, up to 20.000 10-6 weight ratio, was studied. The strong affinity of these nitrides for the oxygen avoids to synthesize pure compounds (no oxygen) by direct combination of the elements or from hydride. The process expected to be used in nuclear fuel industry was chosen. The nitride was prepared by carboreduction and nitridation of the oxide, then ground with different amounts of oxide. The powder obtained was cold pressed and sintered (T = 17200C - 18000C ; t > 15 hours). Analysis of carbon and oxygen content, X ray diffraction measurements, ceramography and electronprobe microanalysis were used to characterize the pellets. The main results are: The oxide (UO2 or MO2) forms at temperatures higher than about 11500C, an oxinitride in contact with nitride matrix (UN or MN), only under nitrogen. This oxinitride, isomorphous with UO2 crystal, is stable up to 17500C with nitride matrix, under a pressure of 1 bar. During the cooling the oxinitride is decomposed in UO2 and U2N3+x. This mixed oxinitride of U and Pu was observed for the first time. The plutonium content of this solid solution is twice smaller than in the nitride matrix. The solubility limit of oxygen in the UN and U0.8Pu0.2N is less than 1000.10-6 weight ratio. This value is lower than published results. The lattice parameter of UN increases in ratio with carbon content, but no noticeable influence of oxygen was detected. This lattice parameter, for UN saturated with oxygen, is 0.48887 ± 5.10-5 nm

  2. The Electrodiffusion of Trace Elements in γ-Cerium, γ-Uranium and ϵ-Plutonium

    International Nuclear Information System (INIS)

    Measurements of solid state diffusion, in an electric field, of various metals in trace concentrations have been made using cerium, uranium and plutonium as solvent metals. An apparatus is described which permits sustained experiments in a controlled atmosphere under constant temperature conditions. Extensive data have been obtained in the case of cerium in the temperature range of 490 - 650°C at current densities from 250 to 500 A/cm2 and over times up to 240 hours. Data are presented for a dozen solute elements. In the case of some transition elements, notably iron, cobalt and nickel, the migration is quite rapid. The use of radioactive tracers, where possible, provided data for quantitative treatment of the results. Spectroscopic analysis provided additional information. Migration rates in uranium measured at 900°C were lower and reduced even more in plutonium at 500°C. However, it was still possible to measure a rate of electrodiffusion of iron. No movement was detected for antimony, magnesium, manganese, silicon or zirconium. With the exception of molybdenum and tin, the metals studied migrated towards the anode. Electrodiffusion presumably results from the net effect of the electric field acting on the ions and from momentum interchange between the ions and conduction electrons. The two effects may thus oppose or reinforce each other. It is felt that the field effect is greater in the cases studied. The tendency of the solute element to form a compound with the solvent metal is one measure of whether migration is to be expected. It is also shown that a relative size effect is important. An interesting aspect of the electrodiffusion of iron in cerium is the very low (∼2 kcal) activation energy. Some comparisons have been made with chemical gradient diffusion. (author)

  3. Study on fuel failure behaviour of plutonium-uranium mixed oxide fuel with NSRR, (1)

    International Nuclear Information System (INIS)

    A research programme is being planned to examine fuel behaviour of plutonium-uranium mixed oxide fuel in water under reactivity initiated accident conditions with Nuclear Safety Research Reactor (NSRR). The plutonium content of the mixed oxide fuel is small as in the case of mixed oxide fuel for thermal reactors. In preparation for the programme, the following had been carried out: design and fabrication in trial of a capsule, evaluations of the energy deposition in a test fuel rod and reactivity worth of a capsule, preliminary experiment with UO2 fuel, and fabrication of mixed oxide fuel pellets. Safety evaluation by the Government of the programme and making of a fuel transportation cask are now in progress. Described in this report are plans in detail of the programme, results of the nuclear characteristic evaluation and preliminary experiment, and the development of a capsule and a transportation cask, safety evaluation by a JAERI ad hoc committee and data in this connection. Preliminary experiments showed that an energy deposition of 300 cal/g. fuel could be attained in a doubly-encapsulated 6.33 wt.% PuO2-UO2 fuel rod by a single pulse irradiation in NSRR; the capsule was also adequate. (author)

  4. Plutonium and uranium contamination in soils from former nuclear weapon test sites in Australia

    Energy Technology Data Exchange (ETDEWEB)

    Child, D.P., E-mail: dpc@ansto.gov.au [Australian Nuclear Science and Technology Organisation, Locked Bag 2001, Kirrawee DC, NSW 2232 (Australia); Department of Earth and Planetary Sciences, Macquarie University, NSW 2109 (Australia); Hotchkis, M.A.C. [Australian Nuclear Science and Technology Organisation, Locked Bag 2001, Kirrawee DC, NSW 2232 (Australia)

    2013-01-15

    The British government performed a number of nuclear weapon tests on Australian territory from 1952 through to 1963 with the cooperation of Australian government. Nine fission bombs were detonated in South Australia at Emu Junction and Maralinga, and a further three fission weapons were detonated in the Monte Bello Islands off the coast of Western Australia. A number of soil samples were collected by Australian Radiation Laboratories in 1972 and 1978 during field surveys at these nuclear weapon test sites. They were analysed by gamma spectrometry and, for a select few samples, by alpha spectrometry to measure the remaining activities of fission products, activation products and weapon materials. We have remeasured a number of these Montebello Islands and Emu Junction soil samples using the ANTARES AMS facility, ANSTO. These samples were analysed for plutonium and uranium isotopic ratios and isotopic concentrations. Very low {sup 240}Pu/{sup 239}Pu ratios were measured at both sites ({approx}0.05 for Alpha Island and {approx}0.02 for Emu Field), substantially below global fallout averages. Well correlated but widely varying {sup 236}U and plutonium concentrations were measured across both sites, but {sup 233}U did not correlate with these other isotopes and instead showed correlation with distance from ground zero, indicating in situ production in the soils.

  5. Plutonium and uranium contamination in soils from former nuclear weapon test sites in Australia

    International Nuclear Information System (INIS)

    The British government performed a number of nuclear weapon tests on Australian territory from 1952 through to 1963 with the cooperation of Australian government. Nine fission bombs were detonated in South Australia at Emu Junction and Maralinga, and a further three fission weapons were detonated in the Monte Bello Islands off the coast of Western Australia. A number of soil samples were collected by Australian Radiation Laboratories in 1972 and 1978 during field surveys at these nuclear weapon test sites. They were analysed by gamma spectrometry and, for a select few samples, by alpha spectrometry to measure the remaining activities of fission products, activation products and weapon materials. We have remeasured a number of these Montebello Islands and Emu Junction soil samples using the ANTARES AMS facility, ANSTO. These samples were analysed for plutonium and uranium isotopic ratios and isotopic concentrations. Very low 240Pu/239Pu ratios were measured at both sites (∼0.05 for Alpha Island and ∼0.02 for Emu Field), substantially below global fallout averages. Well correlated but widely varying 236U and plutonium concentrations were measured across both sites, but 233U did not correlate with these other isotopes and instead showed correlation with distance from ground zero, indicating in situ production in the soils.

  6. Destructive analysis capabilities for plutonium and uranium characterization at Los Alamos National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Tandon, Lav [Los Alamos National Laboratory; Kuhn, Kevin J [Los Alamos National Laboratory; Drake, Lawrence R [Los Alamos National Laboratory; Decker, Diana L [Los Alamos National Laboratory; Walker, Laurie F [Los Alamos National Laboratory; Colletti, Lisa M [Los Alamos National Laboratory; Spencer, Khalil J [Los Alamos National Laboratory; Peterson, Dominic S [Los Alamos National Laboratory; Herrera, Jaclyn A [Los Alamos National Laboratory; Wong, Amy S [Los Alamos National Laboratory

    2010-01-01

    Los Alamos National Laboratory's (LANL) Actinide Analytical Chemistry (AAC) group has been in existence since the Manhattan Project. It maintains a complete set of analytical capabilities for performing complete characterization (elemental assay, isotopic, metallic and non metallic trace impurities) of uranium and plutonium samples in different forms. For a majority of the customers there are strong quality assurance (QA) and quality control (QC) objectives including highest accuracy and precision with well defined uncertainties associated with the analytical results. Los Alamos participates in various international and national programs such as the Plutonium Metal Exchange Program, New Brunswick Laboratory's (NBL' s) Safeguards Measurement Evaluation Program (SME) and several other inter-laboratory round robin exercises to monitor and evaluate the data quality generated by AAC. These programs also provide independent verification of analytical measurement capabilities, and allow any technical problems with analytical measurements to be identified and corrected. This presentation will focus on key analytical capabilities for destructive analysis in AAC and also comparative data between LANL and peer groups for Pu assay and isotopic analysis.

  7. Destructive analysis capabilities for plutonium and uranium characterization at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Los Alamos National Laboratory's (LANL) Actinide Analytical Chemistry (AAC) group has been in existence since the Manhattan Project. It maintains a complete set of analytical capabilities for performing complete characterization (elemental assay, isotopic, metallic and non metallic trace impurities) of uranium and plutonium samples in different forms. For a majority of the customers there are strong quality assurance (QA) and quality control (QC) objectives including highest accuracy and precision with well defined uncertainties associated with the analytical results. Los Alamos participates in various international and national programs such as the Plutonium Metal Exchange Program, New Brunswick Laboratory's (NBL' s) Safeguards Measurement Evaluation Program (SME) and several other inter-laboratory round robin exercises to monitor and evaluate the data quality generated by AAC. These programs also provide independent verification of analytical measurement capabilities, and allow any technical problems with analytical measurements to be identified and corrected. This presentation will focus on key analytical capabilities for destructive analysis in AAC and also comparative data between LANL and peer groups for Pu assay and isotopic analysis.

  8. Reduction of plutonium(IV) using photochemically generated uranium(IV)

    International Nuclear Information System (INIS)

    The reduction of Pu(IV) using photochemically generated U(IV) has been evaluated as a procedure for possible inclusion in the Savannah River Plant's nuclear fuel reprocessing facility. The ''Purex 2nd Uranium Cycle'' with feed conditions of 400 g/L U, 1 M HNO3, and 10-8 to 10-6 M Pu was identified previously as the most promising stage for application of a photochemical method. Laboratory tests were conducted under similar conditions to determine if the plutonium could be successfully reduced and separated in a two-phase flowing system. The laboratory scale tests, which used primarily a 0.01 M butanol/0.01 M hydrazine reductant combination, demonstrated that reductive stripping of Pu(IV) using photochemically generated U(IV) is a practical method for removing plutonium from the organic phase. The kinetics of the reductive stripping procedure were found to be determined by the mixing rates of the organic and aqueous phases in this simple laboratory-scale system

  9. Coulometry for the determination of uranium and plutonium: past and present

    International Nuclear Information System (INIS)

    Precise and accurate determination of uranium (U) and plutonium (Pu) in nuclear fuels is an essential requirement in nuclear fuel cycle for chemical quality assurance of these materials. The redox based electroanalytical methods viz. Potentiometry and Biamperometry are capable of meeting the requirements of high accuracy and precision using milligram amounts of the analyte. However, use of chemical reagents to perform redox reactions in these methods, generates radioactive liquid waste which needs to be processed to recover plutonium. The analytical waste generated by the controlled-potential coulometric (CPC) method is clean as the change in the oxidation state of the analyte is done by the electrolytic reaction. Therefore, the determination of U and Pu in nuclear fuel materials by controlled-potential coulometry is an attractive option instead of biamperometry and potentiometry. In this report, the work carried out to develop CPC employing indigenous coulometers is discussed. The coulometric results for both U and Pu using indigenous coulometers agreed within ± 0.2% with the biamperometric values. The results indicate that indigenous coulometers are suitable for U and Pu determination in nuclear fuel materials and the CPC method can be employed for nuclear fuel samples. (author)

  10. Analysis of BWR lattices to recycle americium

    International Nuclear Information System (INIS)

    This study was carried out to assess the ability to eliminate meaningful quantities of americium in a primarily thermal neutron flux by 'spiking' modern BWR fuel with this minor actinide (MA). The studies carried out so far include the simulation of modern 10 x 10 BWR lattices employing the Westinghouse lattice physics code PHOENIX-4 alongside validation studies using MCNP5 models of the same lattices that were spatially depleted via the MONTEBURNS code coupling to ORIGEN. When considering the total inventory of minor actinides in Am-spiked pins, excluding isotopes of uranium and plutonium, the results indicate that a reduction of approximately 50% or more in the total mass inventory of these minor actinides is viable within the selected pins. Therefore, these preliminary results have encouraged the extension of this work to the development of improved lattice designs to help optimize the transmutation rates as well as absolute MA inventory reductions. The ultimate goal being to design batches of these advanced BWR bundles alongside multi-cycle core reload strategies. (authors)

  11. Multiconfigurational nature of 5f orbitals in uranium and plutonium and their intermetallic compounds

    Science.gov (United States)

    Booth, Corwin

    2013-03-01

    The structural, electronic, and magnetic properties of U and Pu elements and intermetallics remain poorly understood despite decades of effort, and currently represent an important scientific frontier toward understanding matter. The last decade has seen great progress both due to the discovery of superconductivity in PuCoGa5 and advances in theory that finally can explain fundamental ground state properties in elemental plutonium, such as the phonon dispersion curve, the non-magnetic ground state, and the volume difference between the α and δ phases. A new feature of the recent calculations is the presence not only of intermediate valence of the Pu 5f electrons, but of multiconfigurational ground states, where the different properties of the α and δ phases are primarily governed by the different relative weights of the 5f4, 5f5, and 5f6 electronic configurations. The usual method for measuring multiconfigurational states in the lanthanides is to measure the lanthanide LIII-edge x-ray absorption near-edge structure (XANES), a method that is severely limited for the actinides because the spectroscopic features are not well enough separated. Advances in resonant x-ray emission spectroscopy (RXES) have now allowed for spectra with sufficient resolution to resolve individual resonances associated with the various actinide valence states. Utilizing a new spectrometer at the Stanford Synchrotron Radiation Lightsource (SSRL), RXES data have been collected that show, for the first time, spectroscopic signatures of each of these configurations and their relative changes in various uranium and plutonium intermetallic compounds. In combination with conventional XANES spectra on related compounds, these data indicate such states may be ubiquitous in uranium and plutonium intermetallics, providing a new framework toward understanding properties ranging from heavy fermion behavior, superconductivity, and intermediate valence to mechanical and fundamental bonding behavior in

  12. Communication received from France concerning its policies regarding the management of plutonium. Statements on the management of plutonium and of highly enriched uranium

    International Nuclear Information System (INIS)

    The Director General has received a Note Verbale, dated 2 September 2003, from the Permanent Mission of France to the IAEA in the enclosures of which the Government of France, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2002. The Government of France has also made available a statement of its annual figures for holdings of civil high-enriched uranium (HEU) as of 31 December 2002. In light of the request expressed by the Government of France in its Note Verbale of 28 November 1997 concerning its policies regarding the management of plutonium (INFCIRC/549 of 16 March 1998), the enclosures of the Note Verbale of 2 September 2003 are attached for the information of all Member States

  13. Communication received from France concerning its policies regarding the management of plutonium. Statements on the management of plutonium and of high enriched uranium

    International Nuclear Information System (INIS)

    The Director General has received a Note Verbale dated 27 September 2005 from the Permanent Mission of France to the IAEA in the enclosures of which the Government of France, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2003. The Government of France has also made available a statement of its annual figures for holdings of civil high-enriched uranium (HEU) as of 31 December 2004. In light of the request expressed by the Government of France in its Note Verbale of 28 November 1997 concerning its policies regarding the management of plutonium (INFCIRC/549 of 16 March 1998), the enclosures of the Note Verbale of 27 September 2005 are attached for the information of all Member States

  14. Weapons-grade plutonium dispositioning. Volume 3: A new reactor concept without uranium or thorium for burning weapons-grade plutonium

    International Nuclear Information System (INIS)

    The National Academy of Sciences (NAS) requested that the Idaho National Engineering Laboratory (INEL) examine concepts that focus only on the destruction of 50,000 kg of weapons-grade plutonium. A concept has been developed by the INEL for a low-temperature, low-pressure, low-power density, low-coolant-flow-rate light water reactor that destroys plutonium quickly without using uranium or thorium. This concept is very safe and could be designed, constructed, and operated in a reasonable time frame. This concept does not produce electricity. Not considering other missions frees the design from the paradigms and constraints used by proponents of other dispositioning concepts. The plutonium destruction design goal is most easily achievable with a large, moderate power reactor that operates at a significantly lower thermal power density than is appropriate for reactors with multiple design goals. This volume presents the assumptions and requirements, a reactor concept overview, and a list of recommendations. The appendices contain detailed discussions on plutonium dispositioning, self-protection, fuel types, neutronics, thermal hydraulics, off-site radiation releases, and economics

  15. Communication received from Germany concerning its policies regarding the management of plutonium. Statements on the management of plutonium and of high enriched uranium

    International Nuclear Information System (INIS)

    The Director General has received a letter dated 18 April 2005 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2004. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of high enriched uranium (HEU) as of 31 December 2004. In light of the request expressed by the Federal Republic of Germany in its Note Verbale of 1 December 1997 concerning its policies regarding the management of plutonium (INFCIRC/549 of 16 March 1998), the enclosures of the letter of 18 April 2005 are attached for the information of all Member States

  16. Weapons-grade plutonium dispositioning. Volume 3: A new reactor concept without uranium or thorium for burning weapons-grade plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M.; Schnitzler, B.G.; Fletcher, C.D. [and others

    1993-06-01

    The National Academy of Sciences (NAS) requested that the Idaho National Engineering Laboratory (INEL) examine concepts that focus only on the destruction of 50,000 kg of weapons-grade plutonium. A concept has been developed by the INEL for a low-temperature, low-pressure, low-power density, low-coolant-flow-rate light water reactor that destroys plutonium quickly without using uranium or thorium. This concept is very safe and could be designed, constructed, and operated in a reasonable time frame. This concept does not produce electricity. Not considering other missions frees the design from the paradigms and constraints used by proponents of other dispositioning concepts. The plutonium destruction design goal is most easily achievable with a large, moderate power reactor that operates at a significantly lower thermal power density than is appropriate for reactors with multiple design goals. This volume presents the assumptions and requirements, a reactor concept overview, and a list of recommendations. The appendices contain detailed discussions on plutonium dispositioning, self-protection, fuel types, neutronics, thermal hydraulics, off-site radiation releases, and economics.

  17. Plutonium titration by controlled potential coulometry

    International Nuclear Information System (INIS)

    The LAMMAN (Nuclear Materials Metrology Laboratory) is the support laboratory of the CETAMA (Analytical Method Committee), whose two main activities are developing analytic methods, and making and characterizing reference materials. The LAMMAN chose to develop the controlled potential coulometry because it is a very accurate analytical technique which allows the connection between the quantity of element electrolysed to the quantity of electricity measured thanks to the Faraday's law: it does not require the use of a chemical standard. This method was first used for the plutonium titration and was developed in the Materials Analysis and Metrology Laboratory (LAMM), for upgrading its performances and developing it to the titration of other actinides. The equipment and the material used were developed to allow the work in confined atmosphere (in a glove box), with all the restrictions involved. Plutonium standard solutions are used to qualify the method, and in particular to do titrations with an uncertainty better than 0.1 %. The present study allowed making a bibliographic research about controlled potential coulometry applied to the actinides (plutonium, uranium, neptunium, americium and curium). A full procedure was written to set all the steps of plutonium titration, from the preparation of samples to equipments storage. A method validation was done to check the full procedure, and the experimental conditions: working range, uncertainty, performance... Coulometric titration of the plutonium from pure solution (without interfering elements) was developed to the coulometric titration of the plutonium in presence of uranium, which allows to do accurate analyses for the analyses of some parts of the reprocessing of the spent nuclear fuel. The possibility of developing this method to other actinides than plutonium was highlighted thanks to voltammetric studies, like the coulometric titration of uranium with a working carbon electrode in sulphuric medium. (author)

  18. Plutonium and uranium isotopic analysis: recent developments of the MGA++ code suite

    Energy Technology Data Exchange (ETDEWEB)

    Buckley, W; Clark, D; Parker, W E; Romine, W; Ruhter, W; Wang, T F

    1999-09-17

    The Lawrence Livermore National Laboratory develops sophisticated gamma-ray analysis codes for isotopic determinations of nuclear materials based on the principles of the MultiGroup Analysis (MGA). MGA methodology has been upgraded and expanded and is now comprised of a suite of codes known as MGA++. A graphical user interface has also been developed for viewing the data and the fitting procedure. The code suite provides plutonium and uranium isotopic analysis for data collected with high-purity germanium planar and/or coaxial detector systems. The most recent addition to the MGA++ code suite, MGAHI, analyzes Pu data using higher-energy gamma rays (200 keV and higher) and is particularly useful for Pu samples that are enclosed in thick-walled containers. Additionally, the code suite can perform isotopic analysis of uranium spectra collected with cadmium-zinc-telluride (CZT) detectors. We are currently developing new codes with will integrate into the MGA++ suite. These will include Pu isotopic analysis capabilities for data collected with CZT detectors, and U isotopic analysis with high-purity germanium detectors, which utilizes only higher energy gamma rays. Future development of MGA++ will include a capability for isotopic analyses on mixtures of Pu and U.

  19. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

    International Nuclear Information System (INIS)

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement

  20. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

    Science.gov (United States)

    Marshalkin, V. E.; Povyshev, V. M.

    2015-12-01

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium-uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement.

  1. Bio-sorption of uranium and plutonium with Eichhornia crassipes (Water Hyacinth)

    International Nuclear Information System (INIS)

    The continuous expansion in nuclear energy program and an aim of zero discharge makes waste management a challenging task. Waste effluents containing long-lived radionuclides such as 137Cs, 90Sr, 238+239+240Pu and uranium along with other toxic elements have to he suitably treated to bring down the radioactivity levels before it is discharged in to the environment. Biological materials have emerged as an economic and eco-friendly option for removal of toxic heavy metals to an environmentally safe level. Bio-sorption is a phenomenon of rapid passive metal uptake, an ideal alternative for decontamination of metal containing effluents. Bio-sorption of uranium and plutonium from aqueous solutions by dried biomass of Eichhornia crassipes or water hyacinth, a hyper-accumulator, which can tolerate highly toxic condition, was studied. The adsorption of Pu by roots biomass was seen to be more in the pH range from 3-8 and a similar trend was shown by leaves. The adsorption of U by both roots and leaves was more in the pH range of 4-8. Distribution coefficient for Pu in roots and leaf was an average of 1349 ml/g and 3152 ml/g for uranium studied using a wide activity range from 10 Bq to 200 Bq. The presence of anions inhibited the uptake and showed the trend sulphate> nitrate> chloride>> carbonates. The effect of other cations on the absorption capacity was also checked. Effluent solutions from an effluent treatment plant were also subjected to remediation with this biomass. Biomass related metal removal processes may not necessarily replace existing treatment processes but may complement them. (author)

  2. Analytical control of reducing agents on uranium/plutonium partitioning at purex process

    International Nuclear Information System (INIS)

    Spectrophotometric methods for uranium (IV), hydrazine (N2H4) and its decomposition product hydrazoic acid(HN3), and hydroxylamine (NH2 OH) determinations were developed aiming their applications for the process control of CELESTE I installation at IPEN/CNEN-SP. These compounds are normally present in the U/Pu partitioning phase of the spent nuclear treatment via PUREX process. The direct spectrophotometry was used for uranium (IV) analysis in nitric acid-hydrazine solutions based on the absorption measurement at 648 nm. The azomethine compound formed by reaction of hydrazine and p-dimethylamine benzaldehyde with maximum absorption at 457 nm was the basis for the specific analytical method for hydrazine determination. The hydrazoic acid analysis was performed indirectly by its conversion into ferric azide complex with maximum absorption at 465 nm. The hydroxylamine detection was accomplished based on its selective oxidation to nitrous acid which is easily analyzed by the reaction with Griess reagent. The resulted azocompound gas a maximum absorption at 520 nm. The sensibility of 1,4x10-6M for U(IV) with 0,8% of precision, 1,6x10-6M for hydrazine with 0,8% of precision, 2,3x10-6M hydrazoic acid with 0,9% of precision and 2,5x10-6M for hydroxylamine with 0,8% of precision were achieved. The interference studies have shown that each reducing agent can be determined in the presence of each other without any interference. Uranium(VI) and plutonium have also shown no interference in these analysis. The established methods were adapted to run inside glove-boxes by using an optical fiber colorimetry and applied to process control of the CELESTE I installation. The results pointed out that the methods are reliable and safety in order to provide just-in-time information about process conditions. (author)

  3. Solubility and speciation studies of waste radionuclides pertinent to geologic disposal at Yucca Mountain: Results on neptunium, plutonium and americium in J-13 groundwater; Letter report (R707): Reporting period, October 1, 1985--September 30, 1987

    Energy Technology Data Exchange (ETDEWEB)

    Nitsche, H.; Standifer, E.M.; Lee, S.C.; Gatti, R.C.; Tucker, D.B.

    1988-01-01

    We have studied the solubilities of neptunium, plutonium, and americium in J-13 groundwater from Yucca Mountain (Nevada) at three temperatures and hydrogen ion concentrations. They are 25{degree}, 60{degree}C, and 90{degree}C and pH 5.9, 7.0, and 8.5. The results for 25{degree}C are from a study which we did during FY 1984. We included these previous results in the tables to give more information on the solubility temperature dependence; they were, however, done at only one pH (7.0). The solubilities were studied from oversaturation. The nuclides were added at the beginning of each experiment as NpO{sub 2}{sup +}, Pu{sup 4+}, and Am{sup 3+}. The neptunium solubility decreased with increasing temperature and with increasing pH. The soluble neptunium did not change oxidation state at steady state. The pentavalent neptunium was increasingly complexed by carbonate with increasing pH. All solids were crystalline and contained carbonate, except the solid formed at 90{degree}C and pH 5.9. We identified this solid as crystalline Np{sub 2}P{sub 5}. The 25{degree}C, pH 7 solid was Na{sub 3}NpO{sub 2}(CO{sub 3}){sub 2} {center_dot} nH{sub 2}O. Plutonium concentrations decreased with increasing temperature and showed no trend with pH. Pu(V) and Pu(VI) were the dominant oxidation states in the supernatant solution; as the amount of Pu(V) increased with pH, Pu(VI) decreed. The steady-state solids were mostly amorphous, although some contained a crystalline component. They contained Pu(IV) polymer and unknown carbonates.

  4. Polonium 210Po, uranium (234U, 238U and plutonium (238Pu, 239+240Pu bioaccumulation in marine birds

    Directory of Open Access Journals (Sweden)

    Strumińska-Parulska D. I.

    2013-04-01

    Full Text Available The aim of this work was the determination of 210Po, 234U, 238U, 238Pu and 239+240Pu concentration in marine birds which permanently or temporally live in the southern Baltic Sea coast. We chose 11 species of seabirds: three species permanently residing at southern Baltic Sea, four species of wintering birds and three species of migrating birds. The results show that analyzed radionuclides are non-uniformly distributed in the marine birds. The highest activities of 210Po were observed in feathers, muscles and liver. The highest uranium content was found in liver, rest of viscera and feathers, while plutonium in the digestion organs and feathers. Omnivore seabirds accumulated more polonium, plutonium than species that feed on fish, while herbivore seabirds accumulated more uranium than carnivore.

  5. The U-Pu inspector, a new instrument to determine the isotopic compositions of uranium and plutonium

    International Nuclear Information System (INIS)

    The U/Pu-InSpector is a new integrated, portable instrument that can measure the isotopic composition of samples containing uranium and/or plutonium without prior calibration and without the need for skilled operators. It consists of a Low Energy Germanium detector in a Multi-attitude Cryostat (MAC). A shield and collimator are built-in, directly around the detector element, reducing the weight of this detector and shield to approximately 8 kg with a full dewar. The dewar can quickly and easily be filled with a self-pressurizing funnel. The detector is connected to a small portable battery operated analyzer and a Notebook computer. The spectra are automatically stored and analyzed with the help of the MGA codes for plutonium and/or for uranium. 5 refs., 1 fig

  6. Non-destructive assay system for uranium and plutonium in input dissolver solution of Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    A nondestructive assay system for the accountability of uranium and plutonium in input dissolver solution of a nuclear reprocessing plant, named 'Richman's Densitometer', has been developed at the Tokai Reprocessing Plant (TRP). The development of this system has been carried out as a part of Japan Support Program for Agency Safeguards (JASPAS). The system is divided into two nondestructive assay parts, K-edge densitometer (KED) and X-ray fluorescence (XRF) spectrometer. The K-edge densitometry is used to determine the uranium concentration, whereas XRF analysis is used to determine U/Pu weight ratio. The plutonium concentration can be calculated from both the measurement results. The principal of richman's densitometry and the experimental results are discussed in this paper. (author)

  7. Americium(3) coordination chemistry: An unexplored diversity of structure and bonding

    Energy Technology Data Exchange (ETDEWEB)

    Fedosseev, A.M.; Grigoriev, M.S.; Budantseva, N.A. [A.N. Frumkin Institute of Physical Chemistry and Electrochemistry, Moscow (Russian Federation); Guillaumont, D.; Den Auwer, Ch.; Moisy, Ph. [CEA Marcoule, Nuclear Energy Division, RadioChemistry and Processes Department, 30 (France); Le Naour, C.; Simoni, E. [CNRS, University Paris-11 Orsay, IPN, 91 - Orsay (France)

    2010-06-15

    The comparison of the physicochemical behavior of the actinides with that of the lanthanides can be justified by the analogy of their electronic structure, as each of the series is made up of elements corresponding to the filling of a given (n)f atomic shell. However relatively few points of comparison are available, given the lack of available structure for trans-plutonium(III) elements and the additional difficulty of stabilizing coordination complexes of uranium(III) to plutonium(III). This contribution is a focal point of trans-plutonium(III) chemistry and, more specifically, of some americium compounds that have been recently synthesized, all related with hard acid oxygen donor ligands that may be involved in the reprocessing chain of nuclear fuel. After a brief review of the solid hydrates and aquo species for the lanthanide and actinide families, we discuss two types of ligands that have in common three carboxylic groups, namely the amino-tri-acetic acid and the citric acid anions. The additional roles of the nitrogen atom for the first one and of the hydroxy function for the second one are discussed. Accordingly, five new complexes with either americium or lanthanides elements are described: [Co(NH{sub 3}){sub 6}][M(NTA){sub 2}(H{sub 2}O)].8H{sub 2}O with M Nd, Yb and Am, and [Co(NH{sub 3}){sub 6}]{sub 2}K[M{sub 3}(Cit){sub 4}(H{sub 2}O){sub 3}].18H{sub 2}O with Nd and Am cations. In all cases the americium complexes are isostructural with their lanthanide equivalents. (authors)

  8. Americium(3) coordination chemistry: An unexplored diversity of structure and bonding

    International Nuclear Information System (INIS)

    The comparison of the physicochemical behavior of the actinides with that of the lanthanides can be justified by the analogy of their electronic structure, as each of the series is made up of elements corresponding to the filling of a given (n)f atomic shell. However relatively few points of comparison are available, given the lack of available structure for trans-plutonium(III) elements and the additional difficulty of stabilizing coordination complexes of uranium(III) to plutonium(III). This contribution is a focal point of trans-plutonium(III) chemistry and, more specifically, of some americium compounds that have been recently synthesized, all related with hard acid oxygen donor ligands that may be involved in the reprocessing chain of nuclear fuel. After a brief review of the solid hydrates and aquo species for the lanthanide and actinide families, we discuss two types of ligands that have in common three carboxylic groups, namely the amino-tri-acetic acid and the citric acid anions. The additional roles of the nitrogen atom for the first one and of the hydroxy function for the second one are discussed. Accordingly, five new complexes with either americium or lanthanides elements are described: [Co(NH3)6][M(NTA)2(H2O)].8H2O with M Nd, Yb and Am, and [Co(NH3)6]2K[M3(Cit)4(H2O)3].18H2O with Nd and Am cations. In all cases the americium complexes are isostructural with their lanthanide equivalents. (authors)

  9. Thermodynamics of extraction of plutonium(IV) and americium(III) in polyethylene glycol (PEG) and (NH4)2SO4 based aqueous biphasic system (ABS) using 18-crown-6

    International Nuclear Information System (INIS)

    Extraction of plutonium(IV) and americium(III) using PEG-2000/(NH4)2SO4 (40% w/w of each) aqueous biphasic system (ABS) with 18-crown-6 (18-C-6) was studied at four different temperatures in the range of 288 to 318 K. The species extracted are identified to be [Pu.2(18-C-6)](SO4)2 and [Am.2(18-C-6)](SO4)1.5 for Pu(IV) and Am(III), respectively, by the slope ratio method. The thermodynamic parameters evaluated at 298 K by the temperature coefficient approach show that the reaction is favoured by decrease of enthalpy and counteracted by decrease in entropy in the case of Pu(IV) as well as Am(III). The large decrease in the enthalpy observed indicates that there is direct bonding of crown ether to the central metal atom i.e. the formation of inner sphere complex for both Pu(IV) and Am(III) and is similar to that reported previously for Pu(VI). The order of equilibrium constant K and ΔG value is Pu(IV) > Pu(VI) > Am(III) and this is in accordance with the axial charge experienced by the incoming ligand. (orig.)

  10. X-ray photoemission spectroscopy (XPS) study uranium, neptunium, and plutonium oxides in silicate-based glasses

    International Nuclear Information System (INIS)

    Using XPS as the principal investigative tool, the bonding properties of selected metal oxides added to silicate glass are being examined. In this paper, the results of XPS studies of uranium, plutonium, and neptunium in binary and multicomponent silicate-based glasses are presented. Models are proposed to account for the very diverse bonding properties of 6+ and 4+ actinide ions in glass. 15 references, 3 figures, 1 table

  11. X-ray photoemission spectroscopy (XPS) study of uranium, neptunium and plutonium oxides in silicate-based glasses

    International Nuclear Information System (INIS)

    Using XPS as the principal investigative tool, we are in the process of examining the bonding properties of selected metal oxides added to silicate glass. In this paper, we present results of XPS studies of uranium, neptunium, and plutonium in binary and multicomponent silicate-based glasses. Models are proposed to account for the very diverse bonding properties of 6+ and 4+ actinide ions in the glasses

  12. Functional design criteria for the 242-A evaporator and PUREX [Plutonium-Uranium Extraction] Plant condensate interim retention basin

    International Nuclear Information System (INIS)

    This document contains the functional design criteria for a 26- million-gallon retention basin and 10 million gallons of temporary storage tanks. The basin and tanks will be used to store 242-A Evaporator process condensate, the Plutonium-Uranium Extraction (PUREX) Plant process distillate discharge stream, and the PUREX Plant ammonia scrubber distillate stream. Completion of the project will allow both the 242-A Evaporator and the PUREX Plant to restart. 4 refs

  13. Reference computations of public dose and cancer risk from airborne releases of uranium and Class W plutonium

    International Nuclear Information System (INIS)

    This report presents ''reference'' computations that can be used by safety analysts in the evaluations of the consequences of postulated atmospheric releases of radionuclides from the Rocky Flats Environmental Technology Site. These computations deal specifically with doses and health risks to the public. The radionuclides considered are Class W Plutonium, all classes of Enriched Uranium, and all classes of Depleted Uranium. (The other class of plutonium, Y, was treated in an earlier report.) In each case, one gram of the respirable material is assumed to be released at ground leveL both with and without fire. The resulting doses and health risks can be scaled to whatever amount of release is appropriate for a postulated accident being investigated. The report begins with a summary of the organ-specific stochastic risk factors appropriate for alpha radiation, which poses the main health risk of plutonium and uranium. This is followed by a summary of the atmospheric dispersion factors for unfavorable and typical weather conditions for the calculation of consequences to both the Maximum Offsite Individual and the general population within 80 km (50 miles) of the site

  14. Uncertainty induced by chest wall thickness assessment methods on lung activity estimation for plutonium and americium: a large population-based study

    International Nuclear Information System (INIS)

    In vivo lung counting aims at assessing the retained activity in the lungs. The calibration factor relating the measured counts to the worker’s specific retained lung activity can be obtained by several means and strongly depends on the chest wall thickness. Here we compare, for 374 male nuclear workers, the activity assessed with a reference protocol, where the material equivalent chest wall thickness is known from ultrasound measurements, with two other protocols. The counting system is an array of four germanium detectors. It is found that non site-specific equations for the assessment of the chest wall thickness induce large biases in the assessment of activity. For plutonium isotopes or 241Am the proportion of workers for whom the retained activity is within ± 10% of the reference one is smaller than 10%. The use of site-specific equations raises this proportion to 20% and 58% for plutonium and 241Am, respectively. Finally, for the studied population, when site-specific equations are used for the chest wall thickness, the standard uncertainties for the lung activity are 42% and 12.5%, for plutonium and 241Am, respectively. Due to the relatively large size of the studied population, these values are a relatively robust estimate of the uncertainties due to the assessment of the chest wall thickness for the current practice at this site. (paper)

  15. Off gas processing device for degreasing furnace for uranium/plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    A low melting ingredient capturing-cooling trap connected to a degreasing sintering furnace by way of sealed pipelines, a burning/decomposing device for decomposing high melting ingredient gases discharged from the cooling trap by burning them and a gas sucking means for forming the flow of off gases are contained in a glovebox, the inside pressure of which is kept negative. Since the degreasing sintering furnace for uranium/plutonium mixed oxide fuels is disposed outside of the glovebox, operation can be performed safely without greatly increasing the scale of the device, and the back flow of gases is prevented easily by keeping the pressure in the inside of the glovebox negative. Further, a heater is disposed at the midway of the sealed pipelines from the degreasing sintering furnace to the cooling trap, the temperature is kept high to prevent deposition of low melting ingredients to prevent clogging of the sealed pipelines. Further, a portion of the pipelines is made extensible in the axial direction to eliminate thermal stresses caused by temperature change thereby enabling to extend the life of the sealed pipelines. (N.H.)

  16. Temperature and composition effect on the volumic masses of uranium and plutonium nitric acid solutions

    International Nuclear Information System (INIS)

    With an oscillating U-tube densimeter, we measured at different temperatures the volumic masses of uranium, plutonium and nitric acid solutions. By multiple linear regression analysis we found the representative equation of the different parameters in fluence on the volumic mass. We have obtained from the tabulated data the representative equation of the pure water between 15 and 400C with a better precision than 2.10-5g.cm-3. From our own results we have obtained the equations concerning the nitric acid between 0 and 8 M at 200C, then between 15 and 400C. The agreement with our experimental results and with the tabulated data at 200C is in some 10-4g.cm-3 range. At last the general equation which comprises the above found terms is in concordance with the experimental results in +- 1,2.10-3g.cm-3 bracket in the scanned range (200C 0C - 2 M 3] -1 -1 - 1 gl-1 -1). Nevertheless we are not able to explain the difference of 2,3.10-3 g.cm-3 between the constant term of this equation and the choosed origin (the volumic mass of the pure water at 200C). The variations observed between our results and these determined in other laboratories for two constituents systems may be explain by errors in free acidity measurements where these is still no complete agreement between laboratories

  17. TREATMENT OF PLUTONIUM- AND URANIUM-CONTAMINATED OIL FROM ROCKY FLATS ENVIRONMENTAL TECHNOLOGY SITE

    International Nuclear Information System (INIS)

    A removal method for plutonium and uranium has been tested at the Rocky Flats Environmental Technology Site (RFETS). This alternative treatment technology is applicable to U.S. Department of Energy (DOE) organics (mainly used pump oil) contaminated with actinides. In our studies, greater than 70% removal of the actinides was achieved. The technology is based on contacting the oil with a sorbent powder consisting of a surface modified mesoporous material. The SAMMS (Self-Assembled Monolayers on Mesoporous Support) technology was developed by the Pacific Northwest National Laboratory for removal and stabilization of RCRA (i.e., lead, mercury, cadmium, silver, etc.) and actinides in water and for removal of mercury from organic solvents [1, 2]. The SAMMS material is based on self-assembly of functionalized monolayers on mesoporous oxide surfaces. The unique mesoporous oxide support provides a high surface area, thereby enhancing the metal-loading capacity. The testing described in this report was conducted on a small scale but larger-scale testing of the technology has been performed on mercury-contaminated oil without difficulty [3

  18. Application of microwaves in the denitration of nitric solutions of uranium and/or plutonium

    International Nuclear Information System (INIS)

    A method for the conversion of nitric solutions of uranium and/or plutonium that would be an alternative more economic and operatively simpler than the conventional processes is the direct denitration by means of microwaves and vacuum application. This conversion method has the following technical advantages: a) the process is simple, which allows a stable operation; b) neither the addition of chemical reagents nor the dilution of the starting solution are required, thereby the volume of residual liquids is small as compared with other processes; c) one fraction of the evaporation residues is nitric acid which can be reused. The development (on laboratory scale) of this conversion process was initiated. In this first stage, a description of the employed equipment is presented. An example of one of the evaporation and denitration batches and obtained products are fully described. The operative experience leads to deduce that the equipment is satisfactory, due to the following characteristics: 1) it permits an easy manipulation within the glove boxes; 2) the projections, coming out from the reactor, are retained completely; 3) the microwaves oven and the vacuum pump are effectively protected from the corrosive vapors. It is concluded that the employed experimental device is adequate to obtain the necessary materials for the reduction, pressing and sinterability studies. This equipment is adopted for the integral development of sintered pellets fabrication process. (Author)

  19. Caustic Precipitation of Plutonium and Uranium with Gadolinium as a Neutron Poison

    Energy Technology Data Exchange (ETDEWEB)

    VISSER, ANN E.; BRONIKOWSKI, MICHAEL G.; RUDISILL, TRACY S.

    2005-10-18

    The caustic precipitation of plutonium (Pu) and uranium (U) from Pu and U-containing waste solutions has been investigated to determine whether gadolinium (Gd) could be used as a neutron poison for precipitation with greater than a fissile mass containing both Pu and enriched U. Precipitation experiments were performed using both process solution samples and simulant solutions with a range of 2.6-5.16 g/L U and 0-4.3:1 U:Pu. Analyses were performed on solutions at intermediate pH to determine the partitioning of elements for accident scenarios. When both Pu and U were present in the solution, precipitation began at pH 4.5 and by pH 7, 99% of Pu and U had precipitated. When complete neutralization was achieved at pH > 14 with 1.2 M excess OH{sup -}, greater than 99% of Pu, U, and Gd had precipitated. At pH > 14, the particles sizes were larger and the distribution was a single mode. The ratio of hydrogen:fissile atoms in the precipitate was determined after both settling and centrifuging and indicates that sufficient water was associated with the precipitates to provide the needed neutron moderation for Gd to prevent a criticality in solutions containing up to 4.3:1 U:Pu and up to 5.16 g/L U.

  20. Simulasi MCNP5 dalam Eksperimen Kritikalitas Larutan Plutonium Uranium Nitrat Dengan Reflektor Air dan Polyethelene

    Directory of Open Access Journals (Sweden)

    Dinan A.

    2011-12-01

    Full Text Available Banyak perangkat kritik dibangun untuk memenuhi kebutuhan studi fenomena kecelakaan kritikalitas pada larutan fisil di fasilitas daur bahan bakar nuklir. Salah satu diantaranya adalah perangkat kritik SCAMP. Di perangkat ini dikerjakan eksperimen kritikalitas menggunakan bejana silindris stainless steel berisi larutan plutonium uranium nitrat (Pu ditambah U nitrat. Sebanyak 7 eksperimen didemonstrasikan dengan reflektor air di semua sisi permukaan bejana larutan kecuali di bagian atas bejana. Makalah ini membahas simulasi transport Monte Carlo MCNP5 dalam eksperimen kritikalitas larutan Pu ditambah U nitrat dengan reflektor air dan polyethylene. Simulasi MCNP5 dengan pustaka ENDF/BVI memberikan hasil yang paling dekat dengan data eksperimen terutama pada kasus A untuk varian geometri 4. Dibandingkan pustaka ENDF/BV, perhitungan kritikalitas dengan pustaka ENDF/B-VI memberikan hasil lebih dekat dengan perhitungan MONK dimana bias perhitungannya kurang dari 0,44%, khususnya pada kasus A namun pada kasus B dan C simulasi MCNP5dengan pustaka ENDF/BV memberikan hasil dengan kecenderungan lebih baik dibandingkan pustaka ENDFB/VI dengan bias perhitungan kurang dari 2,67% dan kurang dari 1,13%. Secara keseluruhan dapat disimpulkan bahwa MCNP5 telah menunjukkan reliabilitasnya dalam simulasi kritikalitas larutan Pu ditambah U nitrat.

  1. Investigation of Plutonium and Uranium Precipitation Behavior with Gadolinium as a Neutron Poison

    CERN Document Server

    Visser, A E

    2003-01-01

    The caustic precipitation of plutonium (Pu)-containing solutions has been investigated to determine whether the presence of 3:1 uranium (U):Pu in solutions stored in the H-Canyon Facility at the U.S. Department of Energy's (DOE) Savannah River Site (SRS) would adversely impact the use of gadolinium nitrate (Gd(NO3)3) as a neutron poison. In the past, this disposition strategy has been successfully used to discard solutions containing approximately 100 kg of Pu to the SRS high level waste (HLW) system. In the current experiments, gadolinium (as Gd(NO3)3) was added to samples of a 3:1 U:Pu solution, a surrogate 3 g/L U solution, and a surrogate 3 g/L U with 1 g/L Pu solution. A series of experiments was then performed to observe and characterize the precipitate at selected pH values. Solids formed at pH 4.5 and were found to contain at least 50 percent of the U and 94 percent of the Pu, but only 6 percent of the Gd. As the pH of the solution increased (e.g., pH greater than 14 with 1.2 or 3.6 M sodium hydroxide...

  2. Post irradiation examinations of 84F-10A capsule containing uranium-plutonium mixed carbide fuels

    International Nuclear Information System (INIS)

    Two fuel pins filled with uranium-plutonium mixed carbide pellets having different stoichiometry, (U,Pu)C1.0 and (U,Pu)C1.1, were encapsulated in 84F-10A and irradiated in JMTR up to 3.0%FIMA at a peak linear power of 59kW/m. The capsule cooled for ∼4 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. It was found from the radial cross sections of fuel pins that the helium gap between the pellets and the cladding tube was completely closed. At the central part of the fuel pellets the number of small pores was decreased and the grain growth was observed compared with the outer zone. (U,Pu)C1.1 pellets showed higher fission gas release ratio than (U,Pu)C1.0 pellets because the former had relatively high open porosity. Although slight carburization was observed near the inner surface of cladding tube the interaction did not affect the fuel performance itself. (author)

  3. Preparation and study of the nitrides and mixed carbide-nitrides of uranium and of plutonium

    International Nuclear Information System (INIS)

    A detailed description is given of a simple method for preparing uranium and plutonium nitrides by the direct action of nitrogen under pressure at moderate temperatures (about 400 C) on the partially hydrogenated bulk metal. It is shown that there is complete miscibility between the UN and PuN phases. The variations in the reticular parameters of the samples as a function of temperature and in the presence of oxide have been used to detect and evaluate the solubility of oxygen in the different phases. A study has been made of the sintering of these nitrides as a function of the preparation conditions with or without sintering additives. A favorable but non-reproducible, effect has been found for traces of oxide. The best results were obtained for pure UN at 1600 C (96 per cent theoretical density) on condition that a well defined powder, was used. The criterion used is the integral width of the X-ray diffraction lines. The compounds UN and PuN are completely miscible with the corresponding carbides. This makes it possible to prepare carbide-nitrides of the general formula (U,Pu) (C,N) by solid-phase diffusion, at around 1400 C. The sintering of these carbide-nitrides is similar to that of the carbides if the nitrogen content is low; in particular, nickel is an efficient sintering agent. For high contents, the sintering is similar to that of pure nitrides. (author)

  4. Investigation Of Plutonium And Uranium Uptake Into MCU Solvent And Next Generation Solvent

    International Nuclear Information System (INIS)

    At the request of the Savannah River Remediation (SRR) customer, the Savannah River National Laboratory (SRNL) examined the plutonium (Pu) and uranium (U) uptake into the Next Generation Solvent (NGS) that will be used at the Salt Waste Processing Facility (SWPF). SRNL examined archived samples of solvent used in Extraction-Scrub-Strip (ESS) tests, as well as samples from new tests designed explicitly to examine the Pu and U uptake. Direct radiocounting for Pu and U provided the best results. Using the radiocounting results, we found that in all cases there were <3.41E-12 g Pu/g of NGS and <1.17E-05 g U/g of NGS in multiple samples, even after extended contact times and high aqueous:organic volume phase ratios. These values are conservative as they do not allow for release or removal of the actinides by scrub, strip, or solvent wash processes. The values do not account for extended use or any increase that may occur due to radiolytic damage of the solvent.

  5. Plutonium credit in the low enriched uranium fuel of PARR-1

    Energy Technology Data Exchange (ETDEWEB)

    Tayyab, M. [Pakistan Institute of Nuclear Science and Technology, Nuclear Engineering Division, P. O. Nilore, Islamabad (Pakistan)]. E-mail: mtmalikpk@yahoo.com; Bakhtyar, S. [Pakistan Institute of Nuclear Science and Technology, Nuclear Engineering Division, P. O. Nilore, Islamabad (Pakistan); Hamid, T. [Pakistan Institute of Engineering and Applied Sciences, P. O. Nilore, Islamabad (Pakistan)

    2004-08-01

    The amount of plutonium (Pu) isotopes and the resultant savings of {sup 235}U due to their production were calculated in the low enriched uranium (LEU) fuel, being utilized in Pakistan Research Reactor-1 (PARR-1). Further the importance map and relative importance map for different isotopes of Pu were also determined. Equilibrium PARR-1 core was achieved for these calculations. MTR-PC26 package was used to generate the microscopic cross-sections data for 45 elements including fissile/structural materials and also the fission products. Finite difference reactor core analysis code CITATION was employed for the fuel management analysis and static depletion calculations. The results indicated that PARR-1 core has attained its equilibrium state after eleven cycles with each cycle of duration about forty full power (10 MW) days. Further, the results showed that at the beginning of equilibrium cycle (BOEC) of the PARR-1 core, net reactivity addition due to all isotopes of Pu was 4.86 x 10{sup -3}{delta}k/k. Amount of {sup 235}U equivalent to this value of reactivity was found to be 15.58 {+-} 0.021 g. Plots of importance and relative importance maps predicted higher isotopic concentrations of Pu in the fuel elements located in the vicinity of central water box.

  6. Plutonium credit in the low enriched uranium fuel of PARR-1

    International Nuclear Information System (INIS)

    The amount of plutonium (Pu) isotopes and the resultant savings of 235U due to their production were calculated in the low enriched uranium (LEU) fuel, being utilized in Pakistan Research Reactor-1 (PARR-1). Further the importance map and relative importance map for different isotopes of Pu were also determined. Equilibrium PARR-1 core was achieved for these calculations. MTR-PC26 package was used to generate the microscopic cross-sections data for 45 elements including fissile/structural materials and also the fission products. Finite difference reactor core analysis code CITATION was employed for the fuel management analysis and static depletion calculations. The results indicated that PARR-1 core has attained its equilibrium state after eleven cycles with each cycle of duration about forty full power (10 MW) days. Further, the results showed that at the beginning of equilibrium cycle (BOEC) of the PARR-1 core, net reactivity addition due to all isotopes of Pu was 4.86 x 10-3Δk/k. Amount of 235U equivalent to this value of reactivity was found to be 15.58 ± 0.021 g. Plots of importance and relative importance maps predicted higher isotopic concentrations of Pu in the fuel elements located in the vicinity of central water box

  7. Uranium and plutonium determinations for evaluation of high burnup fuel performance

    International Nuclear Information System (INIS)

    Purpose of this work is to experimentally test computational methods being developed for reactor fuel operation. Described are the analytical techniques used in the determination of uranium and plutonium compositions on PWR fuel that has spanned five power cycles, culminating in 55,000 to 57,000 MWd/T burnup. Analyses have been performed on ten samples excised from selected sections of the fuel rods. Hot cell operations required the separation of fuel from cladding and the comminution of the fuel. These tasks were successfully accomplished using a SpectroMil, a ball pestle impact grinding and blending instrument manufactured by Chemplex Industries, Inc., Eastchester, New York. The fuel was dissolved using strong mineral acids and bomb dissolution techniques. Separation of the fuel from fission products was done by solvent (hexone) extraction. Fuel isotopic compositions and assays were determined by the mass spectrometric isotope dilution (MSID) method using NBS standards SRM-993 and SRM-996. Alpha spectrometry was used to determine the 238Pu composition. Relative correlations of composition with burnup were obtained by gamma-ray spectrometry of selected fission products in the dissolved fuel

  8. The characterization of uranium and plutonium reference materials by the (100-X) route; basic principle and applications

    International Nuclear Information System (INIS)

    Uranium and plutonium reference materials are produced to calibrate reagents used in titrimetric methods or to check instrumental techniques like electrochemical methods. For this purpose, the most commonly prepared reference materials are high-purity uranium metal pieces and uranium dioxide pellets, respectively plutonium metal pieces and plutonium dioxide powders. Two different routes can be followed to certify the element content: a direct or an indirect route. With the direct route the main element is assayed using appropriate redox titration methods. The overall uncertainty on the main element is limited by both the precision of the titration methods used and the uncertainty of the reference materials. The indirect route is applicable only for high purity base material. It requires the measurement of the sum of impurities (X) in the material and the substraction of this sum from hundred per cent (100-X). Since the (100-X) route requires the measurement of a large number of elements, a variety of analytical methods has to be used, and a contribution by different specialized laboratories is requested. The uncertainty of the total of impurities is calculated from the respective uncertainties of the determination of each specific element. If the total of impurities is sufficiently small and even if it is determined with a rather large uncertainty, the (100-X) route leads nevertheless to a low uncertainty on the main element content. The application of the (100-X) route is illustrated for the case of uranium dioxide pellets certified as an European Community Nuclear Reference Material. The (100-X) route is recommended as a very accurate one for high purity material characterization and certification purposes

  9. An improved, computer-based, on-line gamma monitor for plutonium anion exchange process control

    International Nuclear Information System (INIS)

    An improved, low-cost, computer-based system has replaced a previously developed on-line gamma monitor. Both instruments continuously profile uranium, plutonium, and americium in the nitrate anion exchange process used to recover and purify plutonium at the Los Alamos Plutonium Facility. The latest system incorporates a personal computer that provides full-feature multichannel analyzer (MCA) capabilities by means of a single-slot, plug-in integrated circuit board. In addition to controlling all MCA functions, the computer program continuously corrects for gain shift and performs all other data processing functions. This Plutonium Recovery Operations Gamma Ray Energy Spectrometer System (PROGRESS) provides on-line process operational data essential for efficient operation. By identifying abnormal conditions in real time, it allows operators to take corrective actions promptly. The decision-making capability of the computer will be of increasing value as we implement automated process-control functions in the future. 4 refs., 6 figs

  10. Non-destructive assay system for uranium and plutonium in input dissolver solution of Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    A nondestructive assay system for the accountability of uranium and plutonium in input dissolver solution of a nuclear reprocessing plant, named 'Richman's Densitometer', has been developed at the Tokai Reprocessing Plant (TRP). The development of this system has been carried out as a part of Japan Support Program for Agency Safeguards (JASPAS). The system was designed, based upon the information of the Hybrid K-edge/XRF Densitometer presented by Ottmar et al., KfK. The instrument, composed of an X-ray generator, detectors, collimators and flow-type cells, was compactly designed and has been installed in a shielded cell. It has been confirmed that the precision for determining uranium concentration (approx. 180 g/l) by K-edge densitometer is 0.2% for 1000sec. counting, whereas XRF for plutonium (approx. 1.5 g/l) performs 1.7% of precision for 3000sec. counting. Further, the system was improved for obtain within 1% of plutonium measurement precision. (author)

  11. SEPARATION OF PLUTONIUM

    Science.gov (United States)

    Maddock, A.G.; Smith, F.

    1959-08-25

    A method is described for separating plutonium from uranium and fission products by treating a nitrate solution of fission products, uranium, and hexavalent plutonium with a relatively water-insoluble fluoride to adsorb fission products on the fluoride, treating the residual solution with a reducing agent for plutonium to reduce its valence to four and less, treating the reduced plutonium solution with a relatively insoluble fluoride to adsorb the plutonium on the fluoride, removing the solution, and subsequently treating the fluoride with its adsorbed plutonium with a concentrated aqueous solution of at least one of a group consisting of aluminum nitrate, ferric nitrate, and manganous nitrate to remove the plutonium from the fluoride.

  12. Preparation of americium amalgam

    International Nuclear Information System (INIS)

    The authors describe a method for the electrochemical preparation of an americium amalgam from americium dioxide and americium 241 and 243 for use in determining the physicochemical properties of the alloy. Moessbauer spectra were made using neptunium dioxide, in the neptunium 237 form, as an absorber. Results show that electrolysis produces a homogeneous amalgam that gives an unoxidized product on vacuum distillation at 200 degrees C

  13. Americium incineration by recycling in target rods using coated particles

    International Nuclear Information System (INIS)

    This paper proposes a type of target rod based on the use of coated particles, for an efficient incineration of americium in nuclear reactors. The analysis takes advantage of the experience gained in the past from long duration irradiation without damage of coated particles with plutonium oxide kernels. A conservative theoretical evaluation of the gas pressure inside the coated particles at the end of irradiation allows comparing the well known conditions of the plutonium oxide particles which were successfully irradiated to high burn-up, with a preliminary design of americium oxide particles. (authors)

  14. Improved analytical sensitivity for uranium and plutonium in environmental samples: Cavity ion source thermal ionization mass spectrometry

    International Nuclear Information System (INIS)

    Following successful field trials, environmental sampling has played a central role as a routine part of safeguards inspections since early 1996 to verify declared and to detect undeclared activity. The environmental sampling program has brought a new series of analytical challenges, and driven a need for advances in verification technology. Environmental swipe samples are often extremely low in concentration of analyte (ng level or lower), yet the need to analyze these samples accurately and precisely is vital, particularly for the detection of undeclared nuclear activities. Thermal ionization mass spectrometry (TIMS) is the standard method of determining isotope ratios of uranium and plutonium in the environmental sampling program. TIMS analysis typically employs 1-3 filaments to vaporize and ionize the sample, and the ions are mass separated and analyzed using magnetic sector instruments due to their high mass resolution and high ion transmission. However, the ionization efficiency (the ratio of material present to material actually detected) of uranium using a standard TIMS instrument is low (0.2%), even under the best conditions. Increasing ionization efficiency by even a small amount would have a dramatic impact for safeguards applications, allowing both improvements in analytical precision and a significant decrease in the amount of uranium and plutonium required for analysis, increasing the sensitivity of environmental sampling

  15. The United States Transuranium and Uranium Registries. Revision 1, [Annual] report, October 1, 1990--April 1992

    Energy Technology Data Exchange (ETDEWEB)

    Kathren, R.L.

    1992-09-01

    This paper describes the history, organization, activities and recent scientific accomplishments of the United States Transuranium and Uranium Registries. Through voluntary donations of tissue obtained at autopsies, the Registries carry out studies of the concentration, distribution and biokinetics of plutonium in occupationally exposed persons. Findings from tissue analyses from more than 200 autopsies include the following: a greater proportion of the americium intake, as compared with plutonium, was found in the skeleton; the half-time of americium in liver is significantly shorter than that of plutonium; the concentration of actinide in the skeleton is inversely proportional to the calcium and ash content of the bone; only a small percentage of the total skeletal deposition of plutonium is found in the marrow, implying a smaller risk from irradiation of the marrow relative to the bone surfaces; estimates of plutonium body burden made from urinalysis typically exceed those made from autopsy data; pathologists were unable to discriminate between a group of uranium workers and persons without known occupational exposure on the basis of evaluation of microscopic kidney slides; the skeleton is an important long term depot for uranium, and that the fractional uptake by both skeleton and kidney may be greater than indicated by current models. These and other findings and current studies are discussed in depth.

  16. Literature review: Phytoaccumulation of chromium, uranium, and plutonium in plant systems

    Energy Technology Data Exchange (ETDEWEB)

    Hossner, L.R.; Loeppert, R.H.; Newton, R.J. [Texas A& M Univ., College Station, TX (United States); Szaniszlo, P.J. [Univ. of Texas, Austin, TX (United States)

    1998-05-01

    Phytoremediation is an integrated multidisciplinary approach to the cleanup of contaminated soils, which combines the disciplines of plant physiology, soil chemistry, and soil microbiology. Metal hyperaccumulator plants are attracting increasing attention because of their potential application in decontamination of metal-polluted soils. Traditional engineering technologies may be too expensive for the remediation of most sites. Removal of metals from these soils using accumulator plants is the goal of phytoremediation. The emphasis of this review has been placed on chromium (Cr), plutonium (Pu), and uranium (U). With the exception of Cr, these metals and their decay products exhibit two problems, specifically, radiation dose hazards and their chemical toxicity. The radiation hazard introduces the need for special precautions in reclamation beyond that associated with non-radioactive metals. The uptake of beneficial metals by plants occurs predominantly by way of channels, pores, and transporters in the root plasma membrane. Plants characteristically exhibit a remarkable capacity to absorb what they need and exclude what they don`t need. But most vascular plants absorb toxic and heavy metals through their roots to some extent, though to varying degrees, from negligible to substantial. Sometimes absorption occurs because of the chemical similarity between beneficial and toxic metals. Some plants utilize exclusion mechanisms, where there is a reduced uptake by the roots or a restricted transport of the metal from root to shoot. At the other extreme, hyperaccumulator plants absorb and concentrate metals in both roots and shoots. Some plant species endemic to metalliferous soils accumulate metals in percent concentrations in the leaf dry matter.

  17. Uranium and plutonium solution assays by transmission-corrected x-ray fluorescence

    Energy Technology Data Exchange (ETDEWEB)

    Ryon, R W; Ruhter, W D; Rudenko, V; Sirontinin, A; Petrov, A A

    1999-09-08

    We have refined and tested a previously developed x-ray fluorescence analysis technique for uranium and plutonium solutions that compensates for variations in the absorption of the exciting gamma rays and fluorescent x-rays. We use {sup 57}Co to efficiently excite the K lines of the elements, and a mixed {sup 57}Co plus {sup 153}Gd transmission source to correct for variations in absorption. The absorption correction is a unique feature of our technique. It is possible to accurately calibrate the system with a single solution standard. There does not need to be a close match in composition (i.e., absorption) between the standard(s) and solutions to be analyzed. Specially designed equipment incorporates a planar intrinsic germanium detector, excitation and transmission radioisotopes, and specimen holder. The apparatus can be inserted into a rubber glove of a glovebox, keeping the apparatus outside and the solutions inside the glovebox, thereby protecting the user and the equipment from possible contamination. An alternate design may be used in chemical reprocessing plants, providing continuous monitoring, by measuring the trans-actinides through stainless steel piping. This technique has been tested at the Bochvar Research Institute of Inorganic Materials in Moscow for possible use in the Russian complex of nuclear facilities. This is part of a cooperative program between laboratories in the United States and Russia to strengthen systems of nuclear materials protection, control, and accountability (MPC and A). A part of this program is to accurately measure and track inventories of materials, thus the need for good non-destructive analytical techniques such as the one described here.

  18. Sorbent Testing for Solidification of Organic Plutonium/Uranium Extraction Waste - Phase IV

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) is evaluating various sorbents to solidify and immobilize hazardous constituents of the organic fraction of plutonium/uranium extraction (PUREX) process waste at the Savannah River Site (SRS).[5] The purpose of the solidification is to provide a cost-effective alternative to incineration of the waste. Incineration at the Consolidated Incinerator Facility (CIF) at SRS is currently identified as the treatment technology for PUREX waste. However, the CIF is not in operation at this time, so SRS is interested in pursuing alternatives to incineration for treatment of this waste. The DOE Western Environmental Technology Office in Butte, MT was designated as the facility for conducting the sorbent testing and evaluation for the organic PUREX waste surrogate. MSE Technology Applications, Inc. tested and evaluated two clay and two polymer sorbents with the capability of solidifying organic PUREX waste. A surrogate organic PUREX waste recipe was utilized, and sorbents were tested and evaluated at bench-scale, 22-liter (5-gallon) scale, and 242-liter (55-gallon) scale. This paper presents experimental results evaluating four sorbent materials including: Imbiber BeadsTM IMB230301-R, Nochar A610 PetrobondTM, Petroset IITM, and Petroset II GranularTM. Previous work at SRS indicated that these products could solidify organic PUREX waste on a bench scale [1]. The sorbents were evaluated using operational criteria and final wasteform properties. Operational criteria included: sorbent capacity; sorption rate; sorbent handling; and mixing requirements. Final wasteform evaluation properties included: ignitability; thermal stability; offgas generation, leachability tests and volumetric expansion. Bench-scale tests, 22-liter (5-gallon) tests, and initial 242-liter (55-gallon) tests are complete. This paper summarizes the results of the bench-scale, 22-liter (5-gallon) scale, and 242-liter (55-gallon) scale tests performed during FY05 with an aqueous

  19. Subsurface bio-mediated reduction of higher-valent uranium and plutonium

    International Nuclear Information System (INIS)

    Bio-mediated reduction of multivalent actinide contaminants plays an important role in their fate and transport in the subsurface. To initiate the process of extending recent progress in uranium biogeochemistry to plutonium, a side-by-side comparison of the bioreduction of uranyl and plutonyl species was conducted with Shewanella alga BrY, a facultative metal-reducing bacterium that is known to enzymatically reduce uranyl. Uranyl was reduced in our system, consistent with literature reports, but we have noted a strong coupling between abiotic and biotic processes and observe that non-reductive pathways to precipitation typically exist. Additionally, a key role of biogenic Fe2+, which is known to reduce uranyl at low pH, is suggested. In contrast, residual organics, present in biologically active systems, reduce Pu(VI) species to Pu(V) species at near-neutral pH. The predominance of relatively weak complexes of PuO2+ is an important difference in how the uranyl and plutonyl species interacted with S. alga. Pu(V) also led to increased toxicity towards S. alga and is also more easily reduced by microbial activity. Biogenic Fe2+, produced by S. alga when Fe(III) is present as an electron acceptor, also played a key role in understanding redox controls and pathways in this system. Overall, the bioreduction of plutonyl is observed under anaerobic conditions, which favors its immobilization in the subsurface. Understanding the mechanism by which redox control is established in biologically active systems is a key aspect of remediation and immobilization strategies for actinides when they are present as subsurface contaminants

  20. Manufacturing experience for mixed uranium-plutonium carbide fuels for fast breeder test reactor

    International Nuclear Information System (INIS)

    The plutonium rich mixed uranium-plutonium carbide pellets of two compositions, namely (U0.3Pu0.70)C (MK-I) and (U0.45Pu0.55)C (MK-II), are used as the fuel for the Indian Fast Breeder Test Reactor (FBTR) at Kalpakkam. These fuels were developed and are being fabricated and characterized at Bhabha Atomic Research Centre (BARC) and have performed very well with peak burn-up exceeding 155GWd/t. This achievement has been possible through a combination of stringent fuel specifications, quality control during fabrication and inputs obtained from the detailed post irradiation examination of fuel at different stages combined with the modeling of the behaviour of the fuel clad and wrapper materials. The high burn-up and short cooled fuel has also been reprocessed successfully in the reprocessing facility at IGCAR. The fissile material (Pu) recovered from reprocessing has now been used for fabrication of fresh mixed carbide fuel which will be loaded in FBTR in the next reload schedule. Closing the carbide fuel cycle is an important milestone in the fast reactor fuel cycle. Bhabha Atomic Research Centre, Trombay developed the fabrication flow sheet for MK-I and MK-II carbide fuels for FBTR. Since carbide fuel is pyrophoric and susceptible to hydrolysis, the fabrication has to be carried out in high purity nitrogen cover gas in leak tight glove boxes. Moreover, adequate shielding is provided to minimize the personnel exposure. The carbide fuel are made using powder metallurgy route with UO2, PuO2 and graphite as the staring material. The homogeneously mixed oxide and graphite powders are compacted into small tablets at low pressure in order to have handling strength and intimate contact between oxide and graphite particles, and to have sufficient porosities for the easy removal of carbon monoxide. The vacuum and temperature for carbothermic reduction are controlled in order to minimize plutonium losses by vaporization and also to have oxygen, nitrogen, carbon, higher carbide

  1. Recovery of Uranium and Plutonium from Waste Matrices Using Supercritical Fluid Extraction

    OpenAIRE

    Krishnamurthy Sujatha; Kancharlapalli Chinaraga Pitchaiah; Nagarajan Sivaraman; Thandankorai Ganapathi Srinivasan; Polur Ranga Rao Vasudeva Rao

    2012-01-01

    Supercritical fluid extraction (SFE) of plutonium in its nitrate form from actual waste, i.e. plutonium bearing cellulose matrix was demonstrated using 0.1 litre capacity extraction vessel. Complete recovery of plutonium was demonstrated using modified supercritical carbon dioxide (Sc-CO2), i.e. Sc-CO2 containing octylphenyl-N,

  2. Impact of High Burnup Uranium Oxide and Mixed Uranium-Plutonium Oxide Water Reactor Fuel on Spent Fuel Management

    International Nuclear Information System (INIS)

    licensed for maximum burnup and enrichment. Increased decay heat places an additional load on plant cooling systems. Increased neutron activity requires radiometric instruments (used to control criticality) to be recalibrated. Increased alpha activity results in increased heat generation. Increased specific activity in reprocessing phases may result in higher radionuclide discharges to the environment and may result in more HLW. These effects can be managed using blending schemes. As the burnup exceeds a certain level, a new reprocessing facility may be needed. The reprocessing of spent MOX fuel presents additional challenges due to the lower solubility of plutonium. For fuel disposal, higher burnup UOX and MOX fuel means higher source terms of the radionuclides leading to a potentially higher release to the groundwater. Higher heat loads could exceed temperature limits in a repository. This may require significant repository operational changes to accommodate higher burnup UOX and MOX, such as increased repository space (although the reduced volume of higher burnup UOX may counteract the need for additional space), smaller waste containers, longer decay times at the surface prior to loading into the repository, and additional shielding during spent fuel transfer from the transportation cask. If reprocessing is the disposition method of choice the increase in discharge burnup has a significant effect on the isotopic quality of recycled fuel. Increased enrichment of the reprocessed uranium or an increased amount of plutonium in MOX fuel will be required to meet the same burnup target. Increases in shielding may be required for fuel refabrication operations.

  3. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water

    Energy Technology Data Exchange (ETDEWEB)

    Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M. [Russian Federal Nuclear Center All-Russian Research Institute of Experimental Physics (Russian Federation)

    2015-12-15

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D{sub 2}O, H{sub 2}O) is proposed. The method is characterized by efficient breeding of the {sup 233}U isotope and safe reactor operation and is comparatively simple to implement.

  4. Assay of the uranium and plutonium content in process residues and wastes using the correction for sample self attenuation in segmented gamma scanning system

    International Nuclear Information System (INIS)

    A method of the correction for sample self attenuation used in segmented gamma scanner (SGS) is described. The method of calibration for assaying the uranium and plutonium contents of each category of heterogeneous process residues and wastes in nuclear facilities is studied. The effect of variable measurement parameter on measurement results is also studied. The measurement results of SGS assay is compared with that of the destructive assay (DA), which is aimed at evaluating SGS method for assaying uranium and plutonium content in process residues and wastes. The deviation of two assay results is 3.6%. The SGS assay results and DA assay results both are coincided well in error limits. Four category of heterogeneous process residues and wastes in nuclear facilities have assayed successfully in physical inventory. The uncertainty of measurement results for uranium and plutonium content in process residues and wastes is 5% with 68.3% of confidence level

  5. Chemical, mass spectrometric, spectrochemical, nuclear and radiochemical analysis of nuclear-grade plutonium metal

    International Nuclear Information System (INIS)

    These analytical procedures are designed to show whether a given material meets the purchaser's specifications as to plutonium content, effective fissile content, and impurity content. The following procedures are described in detail: dissolution procedure; plutonium by controlled-potential coulometry; plutonium by amperometric titration with iron(II); plutonium by ceric sulfate titration method; uranium by Arsenazo I spectrophotometric method; thorium by thorin spectrophotometric method; iron by 1,10-phenanthroline spectrophotometric method; iron by 2,2'-bipyridyl spectrophotometric method; chloride by the thiocyanate spectrophotometric method; fluoride by distillation-spectrophotometric method; nitrogen by distillation-Nessler reagent spectrophotometric method; carbon by the direct combustion-thermal conductivity method; sulfur by distillation-spectrophotometric method; isotopic composition by mass spectrometry; Americium-241 by extraction and gamma counting; Americium-241 by gamma counting; gamma-emitting fission products, uranium, and thorium by gamma-ray spectroscopy; rare earths by copper spark spectrochemical method; tungsten, niobium (columbium) and tantalum by spectrochemical method; sample preparation for spectrographic analysis for trace impurities

  6. Burn-up behavior of FBR fuels sourced in uranium and plutonium recycled in PWRs and its influence on fuel economy

    International Nuclear Information System (INIS)

    Considering the strategic security of uranium resources, the authors investigated the effectual use of fuels recycled in a current LWR system (1.1 GWe-class PWRs of standard type), where they supposed the uranium fuel remade by re-enrichment of the recovery uranium from PWR spent fuels and the MOX fuel produced with the recovery plutonium as main fissile, and examined three cases of MOX fuels prepared by making choice of these matrixes among (A) the recovery uranium, (B) the depleted uranium as waste from natural uranium enrichment and (C) the depleted one sourced from recovery uranium re-enrichment. The result suggests that the multi-recycle of fuels in the LWR system brings the decline in fuel qualities. Particularly, the re-enrichment of recovery uranium brings the issue of an increase of 236U in remanufactured fuel. Thus, in order to investigate the burn-up of fast reactor fuels sourced from the PWR fuel system, they designed a model core of practical-FBR with reference to a concept of FBR reported by JAEA, using the SRAC numerical system. The burn-up behavior of FBR fuels was analyzed which were sourced in the original uranium spent-fuel and the remade uranium spent-fuel. And also, the breeding behavior of blanket materials was investigated which were individually of the depleted natural uranium, the recovery uranium from the original uranium spent-fuel and the recovery uranium from the remade uranium spent-fuel. The fissile 235U in FBR fuels reduces the burden of plutonium while the containment of 236U declines the neutron multiplication in FBRs. (author)

  7. Determination of Nitric Acid in Aqueous Solution of Uranium and Plutonium Purification Cycle by Near Infrared Spectroscopy

    Institute of Scientific and Technical Information of China (English)

    LI; Ding-ming; WANG; Lin; ZHANG; Li-hua; GONG; Yan-ping; MU; Ling; WU; Ji-zong

    2012-01-01

    <正>The concentration of nitric acid interfered with the distribution of uranium and plutonium in nuclear fuel reprocessing process. So, in the reprocessing process control analysis, the determination of the free acid plays an important role. Traditional laboratory analytical method of free acid needs large size sample and is time-consuming. Hence, development of fast analytical method for free acid has important significance for the reprocessing process control analysis. Near-infrared spectroscopy (NIRS) has been proved to be a powerful analytical tool and used in various fields, it’s seldom, however, used in spent

  8. The IDA-80 measurement evaluation programme on mass spectrometric isotope dilution analysis of uranium and plutonium. Vol. 3

    International Nuclear Information System (INIS)

    The evaluation data derived from the measurement results of the laboratories participating in the IDA-80 programme have been compiled in tables and graphs. They concern a total of more than 2000 determinations of isotope ratios, isotope abundances and concentrations for uranium and plutonium obtained on test materials of industrial origin which contained fission products, and on fission product free synthetic reference solutions. Comparisons are made with data certified by CBNM and NBS, and estimates are given which were calculated by variance analyses for within- and between laboratory variations. (orig.)

  9. Fabrication of uranium-plutonium mixed nitride fuel pins (89F-3A) for second irradiation test at JMTR

    International Nuclear Information System (INIS)

    A couple of fuel pins of uranium-plutonium mixed nitride, which has a potential as an advanced FBR fuel, was fabricated for the 2nd irradiation test at JMTR in order to obtain informations on the fuel behavior and to establish the fuel performance. A ferritic stainless steel-cladded fuel pin was adopted to depress the mechanical interaction of nitride fuels with cladding materials (FCMI). A comparative examination is planned by the combined use of an austenitic stainless steel-cladded pin. The irradiation of these pins started on January 1991 for the goal burnup of about 50 GWd/t. (author)

  10. Evaluation of the characteristics of uranium and plutonium mixed oxide (MOX) fuel having high burnup and high plutonium content

    International Nuclear Information System (INIS)

    Two kinds of MOX fuel irradiation tests, i.e., MOX irradiation test up to high burnup and MOX having high plutonium content irradiation test, have been performed since JFY 2007 for five years in order to establish technical data concerning MOX fuel behavior during irradiation, which shall be needed in safety regulation of MOX fuel with high reliability. The high burnup MOX irradiation test consists of irradiation extension and post irradiation examination (PIE). The activities done in JFY 2010 are PIE at Kjeller (85 Kr gamma spectrometry and fuel rod puncture test), PIE at Cadarache (destructive post irradiation examination (D-PIE) such as density measurement, optical microscope, SEM and EPMA), and PIE data analysis. In the frame of irradiation test of high plutonium enriched MOX fuel programme, MOX fuel rods with about 14wt % Pu content are being irradiated at BR-2 reactor and corresponding PIE is also being done at PIE facility in Belgium. The activities done in JFY 2010 are irradiation extension through the irradiation cycles from BR-2 02/2010 to BR-2 01/2011 and gamma spectrometry measurement which has been performed as intermittent PIE and can give burnup, power distribution and FP gas release rate. (author)

  11. Preferential decorporation of americium by pulmonary administration of DTPA dry powder after inhalation of aged PuO2 containing americium in rats

    International Nuclear Information System (INIS)

    After inhalation of plutonium oxides containing various percentages of americium in rats, we identified an acellular transient pulmonary compartment, the epithelial lining fluid (ELF), in which a fraction of actinide oxides dissolve prior to absorption and subsequent extrapulmonary deposit. Chelation therapy is usually considered to be poorly efficient after inhalation of actinide oxides. However, in the present study, prompt pulmonary administration of diethylenetriaminepentaacetic acid (DTPA) as a dry powder led to a decrease in actinide content in ELF together with a limitation of bone and liver deposits. Because americium is more soluble than plutonium, higher amounts of americium were found in ELF, extrapulmonary tissues and urine. Our results also demonstrated that the higher efficacy of DTPA on americium compared to plutonium in ELF induced a preferential inhibition of extrapulmonary deposit and a greater urinary excretion of americium compared to plutonium. All together, our data justify the use of an early and local DTPA treatment after inhalation of plutonium oxide aerosols in which americium can be in high proportion such as in aged compounds. (authors)

  12. Determination of trace quantity of uranium in end product plutonium oxide samples by isotopic dilution mass spectrometry using an indigenous thermal ionization mass spectrometer

    International Nuclear Information System (INIS)

    Characterization of special nuclear materials is an important step for efficient nuclear material management (NMM) in a fuel reprocessing plant. Determination of trace quantity of impurities present in the end product plutonium oxide is very important in categorizing the product for end-user specifications. An accurate knowledge of uranium quantity helps in many ways to characterize the material. Objective of this present work is to optimize the indigenous thermal ionization mass spectrometer for this work and develop a better and effective separation procedure prior to mass spectrometric analysis for determination of trace quantity of uranium in plutonium product stream oxide samples

  13. Buckling and reaction rate experiments in plutonium/uranium metal fuelled, graphite moderated lattices at temperatures up to 400 deg. C. Part I: Experimental techniques and results

    Energy Technology Data Exchange (ETDEWEB)

    Carter, D.H.; Clarke, W.G.; Gibson, M.; Hobday, R.; Hunt, C.; Marshall, J.; Puckett, B.J.; Symons, C.R.; Wass, T. [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1964-07-15

    This report presents experimental measurements of bucklings, flux fine structure and fission rate distributions in graphite moderated lattices fuelled with plutonium/uranium metal at temperatures up to 400 deg. C in the sub-critical assemblies SCORPIO I and SCORPIO II. The experimental techniques employed are described in some detail. The accuracy of the experimental measurements appears to be adequate for testing methods of calculation being developed for the calculation of reactivity and temperature coefficient of reactivity for power reactors containing plutonium and uranium. (author) 26 refs, 17 tabs, 17 figs

  14. Evaluating bis(2-ethylhexyl) methanediphosphonic acid (H2DEH[MDP]) based polymer ligand film (PLF) for plutonium and uranium extraction

    International Nuclear Information System (INIS)

    This paper describes a new analyte extraction medium called polymer ligand film (PLF) that was developed to rapidly extract radionuclides. PLF is a polymer medium with ligands incorporated in its matrix that selectively and quickly extracts analytes. The main focus of the new technique is to shorten and simplify the procedure for chemically isolating radionuclides for determination through alpha spectroscopy. The PLF system was effective for plutonium and uranium extraction. The PLF was capable of co-extracting or selectively extracting plutonium over uranium depending on the PLF composition. The PLF and electrodeposited samples had similar alpha spectra resolutions. (author)

  15. Reevaluation of the average prompt neutron emission multiplicity (nubar) values from fission of uranium and transuranium nuclides

    International Nuclear Information System (INIS)

    In response to a need of the safeguards community, we have begun an evaluation effort to upgrade the recommended values of the prompt neutron emission multiplicity distribution, P/sub nu/ and its average value, nubar. This paper will report on progress achieved thus far. The evaluation of the uranium, plutonium, americium and curium nuclide's nubar values will be presented. The recommended values will be given and discussed. 61 references

  16. Research of natural resources saving by design studies of Pressurized Light Water Reactors and High Conversion PWR cores with mixed oxide fuels composed of thorium/uranium/plutonium

    International Nuclear Information System (INIS)

    Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3. generation, saving of natural resources is of paramount importance for sustainable nuclear energy production. This study consists in the one hand to design high Conversion Reactors exploiting mixed oxide fuels composed of thorium/uranium/plutonium, and in the other hand, to elaborate multi-recycling strategies of both plutonium and 233U, in order to maximize natural resources economy. This study has two main objectives: first the design of High Conversion PWR (HCPWR) with mixed oxide fuels composed of thorium/uranium/plutonium, and secondly the setting up of multi-recycling strategies of both plutonium and 233U, to better natural resources economy. The approach took place in four stages. Two ways of introducing thorium into PWR have been identified: the first is with low moderator to fuel volume ratios (MR) and ThPuO2 fuel, and the second is with standard or high MR and ThUO2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233U production and low plutonium consumption. This second step came up with two specific concepts, from which multi-recycling strategies have been elaborated. The exclusive production and recycling of 233U inside HCPWR limits the annual economy of natural uranium to approximately 30%. It was brought to light that the strong need in plutonium in the HCPWR dedicated to 233U production is the limiting factor. That is why it was eventually proposed to study how the production of 233U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233U in dedicated PWR relaxes the constraint on plutonium inventories and favours the transition toward a symbiotic reactor fleet composed of both PWR and HCPWR loaded with thorium fuel. This strategy is more adapted and leads to an annual economy of natural uranium of about 65%. (author)

  17. Determination of nitrogen in uranium-plutonium mixed oxide fuel by gas chromatography after fusion in an inert gas atmosphere

    International Nuclear Information System (INIS)

    A gas chromatographic technique has been developed for the determination of nitrogen in uranium-plutonium mixed oxide fuel after fusion in an inert gas atmosphere. When the sample and pure iron powder in a graphite crucible were heated to approximately 2500C by a resistance heating furnace, a large amount of carbon monoxide was evolved with a small amount of nitrogen and hydrogen. A gas chromatograph equipped with a pre-cut system was used for the separation of nitrogen from the carbon monoxide. Nitrogen separated by the gas chromatograph was determined by means of a thermal conductivity detector. Only 100mg of the sample was used, and the analysis requires about 10min. No specific skills for glove-box work are necessary. The relative standard deviation and detection limit (3σ-criterion) were less than 5% and 9μgg-1, respectively. The present method is not only applicable to the analysis of research samples but also to the quality control of uranium-plutonium mixed oxide fuel production lines

  18. Managing proliferation risks from civilian and weapon-grade plutonium and enriched uranium: A comprehensive cut-off convention

    International Nuclear Information System (INIS)

    The problem of weapon-grade fissile materials is closely related to the aim of achieving a nuclear-weapon-free world. Huge amounts of highly enriched uranium have been produced for nuclear weapons. More than 1000 tonnes of plutonium emerged as a by-product of civilian nuclear industry. Separated from spent fuel it is readily usable for nuclear weapons. The worldwide civilian tritium inventory may reach the same size as military stocks about the year 2010. This poses an increasing danger of horizontal nuclear proliferation. Production, stockpiling, trade, processing and uses of weapon-grade materials like Highly enriched uranium, plutonium and tritium promote its geographical spread, enlarge the group of people with the relate know-how and create the danger of diversion of material and the proliferation of knowledge for the purpose of weapons production. Therefore, a fundamental turn away from using weapon-grade materials in scientific and economic applications of nuclear energy is desirable in all countries. Priority should be given to using nuclear fuel cycles which are as proliferation resistant as possible. Without this, the continuation of civil nuclear programs seems to be irresponsible and unjustifiable. The role of the IAEA in export control safeguards related to the above problems is indispensable

  19. Investigation of Plutonium and Uranium Precipitation Behavior with Gadolinium as a Neutron Poison

    International Nuclear Information System (INIS)

    The caustic precipitation of plutonium (Pu)-containing solutions has been investigated to determine whether the presence of 3:1 uranium (U):Pu in solutions stored in the H-Canyon Facility at the U.S. Department of Energy's (DOE) Savannah River Site (SRS) would adversely impact the use of gadolinium nitrate (Gd(NO3)3) as a neutron poison. In the past, this disposition strategy has been successfully used to discard solutions containing approximately 100 kg of Pu to the SRS high level waste (HLW) system. In the current experiments, gadolinium (as Gd(NO3)3) was added to samples of a 3:1 U:Pu solution, a surrogate 3 g/L U solution, and a surrogate 3 g/L U with 1 g/L Pu solution. A series of experiments was then performed to observe and characterize the precipitate at selected pH values. Solids formed at pH 4.5 and were found to contain at least 50 percent of the U and 94 percent of the Pu, but only 6 percent of the Gd. As the pH of the solution increased (e.g., pH greater than 14 with 1.2 or 3.6 M sodium hydroxide (NaOH) excess), the precipitate contained greater than 99 percent of the Pu, U, and Gd. After the pH greater than 14 systems were undisturbed for one week, no significant changes were found in the composition of the solid or supernate for each sample. The solids were characterized by X-ray diffraction (XRD) which found sodium diuranate (Na2U2O7) and gadolinium hydroxide (Gd(OH)3) at pH 14. Thermal gravimetric analysis (TGA) indicated sufficient water molecules were present in the solids to thermalize the neutrons, a requirement for the use of Gd as a neutron poison. Scanning electron microscopy (SEM) was also performed and the accompanying back-scattering electron analysis (BSE) found Pu, U, and Gd compounds in all pH greater than 14 precipitate samples. The rheological properties of the slurries at pH greater than 14 were also investigated by performing precipitate settling rate studies and measuring the viscosity and density of the materials. Based on the

  20. Post irradiation examinations of uranium-plutonium mixed carbide fuels irradiated at low linear power rate

    International Nuclear Information System (INIS)

    Two pins containing uranium-plutonium carbide fuels which are different in stoichiometry, i.e. (U,Pu)C1.0 and (U,Pu)C1.1, were constructed into a capsule, ICF-37H, and were irradiated in JRR-2 up to 1.0 at % burnup at the linear heat rate of 420 W/cm. After being cooled for about one year, the irradiated capsule was transferred to the Reactor Fuel Examination Facility where the non-destructive examinations of the fuel pins in the β-γ cells and the destructive ones in two α-γ inert gas atmosphere cells were carried out. The release rates of fission gas were low enough, 0.44 % from (U,Pu)C1.0 fuel pin and 0.09% from (U,Pu)C1.1 fuel pin, which is reasonable because of the low central temperature of fuel pellets, about 1000 deg C and is estimated that the release is mainly governed by recoil and knock-out mechanisms. Volume swelling of the fuels was observed to be in the range of 1.3 ∼ 1.6 % for carbide fuels below 1000 deg C. Respective open porosities of (U,Pu)C1.0 and (U,Pu)C1.1 fuel were 1.3 % and 0.45 %, being in accordance with the release behavior of fission gas. Metallographic observation of the radial sections of pellets showed the increase of pore size and crystal grain size in the center and middle region of (U,Pu)C1.0 pellets. The chemical interaction between fuel pellets and claddings in the carbide fuels is the penetration of carbon in the fuels to stainless steel tubes. The depth of corrosion layer in inner sides of cladding tubes ranged 10 ∼ 15 μm in the (U,Pu)C1.0 fuel and 15 #approx #25 μm in the (U,Pu)C1.1 fuel, which is correlative with the carbon potential of fuels posibly affecting the amount of carbon penetration. (author)

  1. Americium recovery from reduction residues

    Science.gov (United States)

    Conner, W.V.; Proctor, S.G.

    1973-12-25

    A process for separation and recovery of americium values from container or bomb'' reduction residues comprising dissolving the residues in a suitable acid, adjusting the hydrogen ion concentration to a desired level by adding a base, precipitating the americium as americium oxalate by adding oxalic acid, digesting the solution, separating the precipitate, and thereafter calcining the americium oxalate precipitate to form americium oxide. (Official Gazette)

  2. Determination of the distribution of plutonium concentration in uranium- and plutonium dioxide by X-ray analysis

    International Nuclear Information System (INIS)

    The diffraction profile measured on a (U, Pu)O2-preparation consists of an equipment broadening function and the physical profile whose broadening is caused by the limited size of the coherently scattering domains, the lattice distortion and the Pu concentration distribution. The determination of the line broadening influences - size of the coherently scattering domains, lattice distortion and Pu concentration distribution - calls for knowledge of the instrument broadening function. This instrument function including the parameters of the diffractometer used, such as the horizontal or vertical divergence of X-rays, the width of the emission lines, the finite slit width of the detector diaphragm etc., was determined theoretically. The separation decovolution of the equipment function from the diffraction profile obtained in the experiment was performed in the Fourier space. Starting from the domain size function and the lattice distortion function a method has been developed which allows to determine the mean domain size, the lattice expansion distribution within a domain and the distribution of plutonium concentrations. (orig./HP)

  3. The United States Transuranium and Uranium Registries: 1968-1993

    International Nuclear Information System (INIS)

    The historical development and scientific contributions of the United States Transuranium and Uranium Registries are briefly traced. Further to encourage radiobiology studies and to provide unique materials for research in other areas such as biomarkers including oncogenes, the Registries have established the National Human Radiobiology Tissue Repository, which includes more than 20,000 tissue samples and solutions, histopathology slides, and related materials from persons with internal depositions, including radium dial painters, available to other investigators for collaborative or individual research purposes. Emphasis is given to recent findings of the Registries with respect to biokinetic models for the actinides and the post-mortem analysis of USTUR Case 0246, an individual who suffered a massive intake of 241Am eleven years prior to death from heart disease. Registries studies show that, in addition to skeleton and liver, muscle is an appreciable reservoir for both plutonium and americium. The systemic biokinetics of plutonium and americium are well represented by a three-compartment model that includes skeleton, liver, and muscle, although the biokinetic parameters of these two transuranium elements differ significantly. Initial uptake fractions for plutonium are 0.4, 0.4, and 0.2, with residence half-times of 50, 20 and 10 years, respectively, for skeleton, liver and muscle. For americium, the comparable uptake fractions are 0.45, 0.25, 0.3 with residence half-times of 50, 2.5 and 10 years for the three compartments. (author)

  4. Precise isotope ratio measurements for uranium, thorium and plutonium by quadrupole-based inductively coupled plasma mass spectrometry

    International Nuclear Information System (INIS)

    Precise long-term measurements of uranium and thorium isotope ratios was carried out in 1 μg/L solutions using a quadrupole inductively coupled plasma mass spectrometer (ICP-QMS). The isotopic ratios of uranium (235U/ 238U = 1, 0.02 and 0.00725) were determined using a cross-flow nebulizer (CFN, at solution uptake rate of 1 mL/min) and a low-flow microconcentric nebulizer (MCN, at solution uptake rate of 0.2 mL/min) over 20 h. For 1 μg/L uranium solution (235U/238U = 1) relative external standard deviations (RESDs) of 0.05% and 0.044% using CFN and MCN, respectively, can be achieved. Additional short term isotope ratio measurements using a direct injection high-efficiency nebulizer (DIHEN) of 1 μg/L uranium solution (235U/238U = 1) at a solution uptake rate of 0.1 mL/min yielded an RSD of 0.06-0.08%. The sensitivity of solution introduction by DIHEN for uranium, thorium and plutonium (145 MHz/ppm, 150 MHz/ppm and 177 MHz/ppm, respectively) increased significantly compared to CFN and MCN and the solution uptake rate can be reduced to 1 μL/ min in DIHEN-ICP-MS. Isotope ratio measurements at an ultralow concentration level (e.g. determination of 240Pu/ 239Pu isotope ratio in a 10 ng/L Pu waste solution) were carried out for the characterization of radioactive waste and environmental samples. (orig.)

  5. On chlorization of uranium and plutonium oxides in NaCl-KCl-MgCl2 molten eutectic

    International Nuclear Information System (INIS)

    The chlorination process of U3O8, UO2, and PuO2 in a melt of anhydrous NaCl-KCl-MgCl2 with gaseous chlorine and carbon tetrachloride has been studied. The chlorination rate of uranium oxides has been studied within a temperature range 500-800 deg C at a chlorine feeding rate of 10 ml/min. Thermoqravimetric and X-ray analyses have shown that K2UO2Cl4 compound is the final product of chlorination of uranium oxides. The mechanism of chlorination has been proposed. THe rate of PuO2 chlorination has been studied within the same temperature range. It has been established that PuO2 is readily chlorinated with CCl4 vapours at a feeding rate of 10 ml/min. In contrast to uranium, chloride forms of plutonium in a highest oxidized state are unstable and are reduced in the melt to Pu(3) and Pu(4). The oxygen being released is retained by CCl4 and by the products of CCl4 pyrolysis

  6. Study of the sintering of hyper-stoichiometric mixed oxides of uranium and plutonium

    International Nuclear Information System (INIS)

    The aim of this work is to study the reactive sintering, in oxidizing conditions, of compounds (U1-yPuy)O2+x, more or less homogeneous, where y is about 0.06 and x, positive, and is inferior to 0.215. It has been shown, at least for the low values of y considered, that it is possible to anticipate the properties influencing the sintering of the oxides, with as support those relative to the sur-stoichiometric uranium oxides. The use of gaseous mixtures presenting an adjustable oxygen potential, is an interesting solution to adjust x to the stoichiometry of the oxides. Either the thermodynamic equilibrium is not reached in the used experimental conditions, the samples, sensitive to the characteristics of the scanning gas, can then evolved by loss or gain of oxygen. The sintering of the considered oxides is carried out at a temperature all the more lower, and with a kinetics all the more high, that the effective ratio of composition O/(U+Pu) is high. Nevertheless, when the ratio to stoichiometry is important (2.215), the presence of the oxide M4O9-z in high proportion compared with the oxide MO2+x during all the thermal treatment cycle slows it considerably. These results are in a direct correlation with the activation energy values of sintering, measured by the temperature methods and constant heating velocities. Indeed, it decreases of 560 kJ.mol-1, for O/M ratios between 1.98 and 2+ε (ε∼0.001), the most probable value for the stoichiometric oxide being of 465 kJ.mol-1; it is constant and equal to 430 kJ.mol-1 when x passes from 0.001 to 0.16; then it increases a lot and in continuous for compositions such as the oxide M4O9-z is majority compared with MO2+ and approaches 640 kJ.mol-1 for a O/M ratio of 2.217. For some samples in the usual oxidizing conditions, whose sintering is ended at about 1100 - 1200 C, a secondary de-densification phenomenon ('solarization') appears if the temperature rising is followed. It is an irreversible process bound to the appearing

  7. Stabilization of mixed carbides of uranium-plutonium by zirconium. Part 1.: uranium carbide with small additions of zirconium; Etude de la stabilisation des carbures mixtes d'uranium et de plutonium par addition de zirconium. 1. partie: etude des carbures d'uranium avec de faibles additions de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Bocker, S. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1969-07-01

    Cast carbide samples, being of a high density and purity, are preferable for research purposes, to samples produced by powder metallurgy methods. Samples of uranium carbide with small additions of zirconium (1 to 5 per cent) were cast, as rods, in an arc furnace. A single phase carbide with interesting qualities was produced. As cast, a dendrite structure is observed, which does not disappear, after a treatment at 1900 deg. C during 110 hours. In comparison with uranium monocarbide the compatibility with stainless steel is much improved. The specific heat (between room temperature and 2500 deg. C) is similar to the specific heat of uranium monocarbide. A study of these mixed carbides, but having a part of the uranium replaced by plutonium is under way. (author) [French] Les echantillons de monocarbures obtenus par coulee sont tres interessants pour les recherches experimentales a cause de leur grande purete, de leur densite tres elevee et de la facilite d'obtention des lingots de forme et dimensions variees. On a prepare et coule dans un four a arc des echantillons de carbures d'uranium avec de faibles additions de zirconium (1 a 5 at. pour cent). On obtient ainsi des carbures monophases presentant de meilleures proprietes que le monocarbure d'uranium. A l'etat brut de coulee on observe une structure dendritique qui n'est pas detruite par un traitement thermique de 110 heures a 1900 deg. C. La compatibilite avec l'acier inoxydable 316 (a 925 deg. C pendant 500 heures) est nettement amelioree par rapport a UC. La chaleur specifique (entre la temperature ordinaire et 2500 deg. C) et la densite sont tres peu differentes de celles du monocarbure d'uranium. Une etude concernant les composes U-Pu-Zr-C est actuellement en cours. (auteur)

  8. Development of a methodology for the determination of americium and thorium by ICP-AES and their inter-element effect

    International Nuclear Information System (INIS)

    Due to the scarcity of good quality uranium resources, the growth of nuclear technology in India is dependent on the utilization of the vast thorium resources. Therefore, Advance Heavy Water Reactor is going to acquire significant role in the scenario of Indian nuclear technology, where (Th, Pu)O2 will be utilized as fuel in the outermost ring of the reactor core. This will lead to a complex matrix containing thorium as well as americium, which is formed due to β-decay of plutonium. The amount of americium is dependent on the burn up and the storage time of the Pu based fuels. In the present case, attempt was made to develop a method for the determination of americium as well as thorium by ICP-AES. Two emission lines of americium were identified and calibration curves were established for determination of americium. Though the detection limit of 283.236 nm line (5 ng mL-1) of americium was found to be better than that of 408.930 nm (11 ng mL-1), the former line is significantly interfered by large amount of thorium. Three analytical lines (i.e. 283.242, 283.730 and 401.913 nm) of thorium were identified and calibration curves were established along with their detection limits. It was observed that 283.242 and 401.913 nm line are having similar detection limits (18 and 13 ng mL-1, respectively) which are better than that of 283.730 nm (60 ng mL-1). This can be attributed to the high background of 283.273 nm channel of thorium. The spectral interference study revealed that even small amount of americium has significant contribution on 283.242 nm channel of thorium while the other two channels remain practically unaffected. Considering both these facts, spectral interference and analytical performance (detection limits and sensitivity), it was concluded that 401.913 nm line is the best analytical line out of the three lines for determination of thorium in presence of americium. (author)

  9. Assessment of uranium, plutonium, and Np-237 content of high level liquid waste on E-Area vault package limits

    International Nuclear Information System (INIS)

    The purpose of this report is to assess the waste tank inventory of uranium, plutonium and Np-237 to determine potential impacts on waste certification for the E-Area vaults (EAV). Procedure WAC 3.10, Rev. 1, of the 1S Manual imposes administrative control limits for radioactive material in waste packages sent to the EAV. Waste tank supernate contains trace amounts of U, Pu, and Np. Thus any material contaminated with supernate and placed in a B-25 waste package may contain one or more of these elements' radioactive isotopes. This report uses material inventory data, solubility data and tank volumes to determine the potential-, for waste packages, contaminated with waste tank supernate, to exceed the administrative control limits of procedure WAC 3. 10, Rev. 1, for U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242

  10. Coordinated safeguards for materials management in a uranium--plutonium nitrate-to-oxide coconversion facility: Coprecal

    International Nuclear Information System (INIS)

    This report describes the conceptual design of an advanced materials-management system for safeguarding special nuclear materials in a uranium--plutonium nitrate-to-oxide coconversion facility based on the Coprecal process. Design concepts are presented for near real-time (dynamic) accountability by forming dynamic materials balances from information provided by chemical and nondestructive analyses and from process-control instrumentation. Modeling and simulation techniques are used to compare the sensitivities of proposed dynamic materials accounting strategies to both abrupt and protracted diversion. The safeguards implications of coconversion as well as some unique features of the reference process are discussed and design criteria are identified to improve the safeguardability of the Coprecal coconversion process

  11. Assessment of uranium, plutonium, and Np-237 content of high level liquid waste on E-Area vault package limits

    Energy Technology Data Exchange (ETDEWEB)

    Clemmons, J.S.

    1994-06-08

    The purpose of this report is to assess the waste tank inventory of uranium, plutonium and Np-237 to determine potential impacts on waste certification for the E-Area vaults (EAV). Procedure WAC 3.10, Rev. 1, of the 1S Manual imposes administrative control limits for radioactive material in waste packages sent to the EAV. Waste tank supernate contains trace amounts of U, Pu, and Np. Thus any material contaminated with supernate and placed in a B-25 waste package may contain one or more of these elements` radioactive isotopes. This report uses material inventory data, solubility data and tank volumes to determine the potential-, for waste packages, contaminated with waste tank supernate, to exceed the administrative control limits of procedure WAC 3. 10, Rev. 1, for U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242.

  12. Preparation of uranium-plutonium carbide-based fuels simulating high burnup by carbothermic reduction and their properties

    International Nuclear Information System (INIS)

    Three types, hypostoichiometric, nearly stoichiometric and hyperstoichiometric, of uranium-plutonium carbide fuels simulating 10 at.% burnup were prepared by carbothermic reduction of oxide containing fission product elements. The carbides contained fission product phases such as the UMoC2 and the U2RuC2 type or the RECsub(1.5-2.0) phases (RE:rare earth). Composite theoretical densities of heterogenious carbides containing the UC, U2C3 type and fission product phases were calculated from the proportions and densities of these phases. By comparison of specific volume of the carbide between of 0 at.% and 10 at.% burnup, the solid fission product swelling rate of a carbide-based fuel was estimated to be 0.4-0.5 % per at.% burnup. (author)

  13. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium

    International Nuclear Information System (INIS)

    We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author)

  14. Analytical techniques for in-line/on-line monitoring of uranium and plutonium in process solutions : a brief literature survey

    International Nuclear Information System (INIS)

    In-line/on-line monitoring of various parameters such as uranium-plutonium-fission product concentration, acidity, density etc. plays an important role in quickly understanding the efficiency of processes in a reprocessing plant. Efforts in studying and installation of such analytical instruments are going on since more than three decades with adaptation of newer methods and technologies. A review on the developement of in-line analytical instrumentation was carried out in this laboratory about two decades ago. This report presents a very short literature survey of the work in the last two decades. The report includes an outline of principles of the main techniques employed in the in-line/on-line monitoring. (author). 77 refs., 6 tabs

  15. High temperature X-ray diffraction study of the oxidation products and kinetics of uranium-plutonium mixed oxides.

    Science.gov (United States)

    Strach, Michal; Belin, Renaud C; Richaud, Jean-Christophe; Rogez, Jacques

    2014-12-15

    The oxidation products and kinetics of two sets of mixed uranium-plutonium dioxides containing 14%, 24%, 35%, 46%, 54%, and 62% plutonium treated in air were studied by means of in situ X-ray diffraction (XRD) from 300 to 1773 K every 100 K. The first set consisted of samples annealed 2 weeks before performing the experiments. The second one consisted of powdered samples that sustained self-irradiation damage. Results were compared with chosen literature data and kinetic models established for UO2. The obtained diffraction patterns were used to determine the temperature of the hexagonal M3O8 (M for metal) phase formation, which was found to increase with Pu content. The maximum observed amount of the hexagonal phase in wt % was found to decrease with Pu addition. We conclude that plutonium stabilizes the cubic phases during oxidation, but the hexagonal phase was observed even for the compositions with 62 mol % Pu. The results indicate that self-irradiation defects have a slight impact on the kinetics of oxidation and the lattice parameter even after the phase transformation. It was concluded that the lattice constant of the high oxygen phase was unaffected by the changes in the overall O/M when it was in equilibrium with small quantities of M3O8. We propose that the observed changes in the high oxygen cubic phase lattice parameter are a result of either cation migration or an increase in the miscibility of oxygen in this phase. The solubility of Pu in the hexagonal phase was estimated to be below 14 mol % even at elevated temperatures. PMID:25412433

  16. Miniaturization of uranium/plutonium/fission products separation: design of a 'lab-on-CD' micro-system and application

    International Nuclear Information System (INIS)

    The chemical analysis of spent nuclear fuels is essential to design future nuclear fuels cycle and reprocessing methods but also for waste management. The analysis cycle consists of several chemical separation steps which are time consuming and difficult to implement due to confinement in glove boxes. It is required that the separation steps be automated and that the volume of radioactive waste generated be reduced. The design of automated, miniaturized and disposable analytical platforms should fulfill these requirements. This project aims to provide an alternative to the first analytical step of the spent fuels analysis: the chromatographic separation of Uranium and Plutonium from the minor actinides and fission products. The goal is to design a miniaturized platform showing analytical performances equivalent to the current process, and to reduce both the exposure of workers through automation, and the volume of waste produced at the end of the analysis cycle. Thus, the separation has been implemented on a disposable plastic micro-system (COC), specifically designed for automation: a lab on a Compact Disk or lab-on-CD. The developed prototype incorporates an anion-exchange monolithic micro-column whose in-situ synthesis as well as surface functionalization have been optimized specifically for the desired separation. The development of an adapted separation protocol was carried out using a simulation tool modeling the elution of the various elements of interest. This tool is able to predict the column geometry (length and cross section) suited to obtain pure fractions of Uranium and Plutonium as a function of the sample composition. Finally, the prototype is able to automatically carry out four separations simultaneously reducing the number of manipulations, the analysis time and reducing the volume of liquid waste by a factor of 1000. (author)

  17. An investigation to compare the performance of methods for the determination of free acid in highly concentrated solutions of plutonium and uranium nitrate

    International Nuclear Information System (INIS)

    An investigation has been carried out to compare the performance of the direct titration method and the indirect mass balance method, for the determination of free acid in highly concentrated solutions of uranium nitrate and plutonium nitrate. The direct titration of free acid with alkali is carried out in a fluoride medium to avoid interference from the hydrolysis of uranium or plutonium, while free acid concentration by the mass balance method is obtained by calculation from the metal concentration, metal valency state, and total nitrate concentration in a sample. The Gran plot end-point prediction technique has been used extensively in the investigation to gain information concerning the hydrolysis of uranium and plutonium in fluoride media and in other complexing media. The use of the Gran plot technique has improved the detection of the end-point of the free acid titration which gives an improvement in the precision of the determination. The experimental results obtained show that there is good agreement between the two methods for the determination of free acidity, and that the precision of the direct titration method in a fluoride medium using the Gran plot technique to detect the end-point is 0.75% (coefficient of variation), for a typical separation plant plutonium nitrate solution. The performance of alternative complexing agents in the direct titration method has been studied and is discussed. (author)

  18. Communication received from the United Kingdom of Great Britain and Northern Ireland concerning its policies regarding the management of plutonium. Statements on the management of plutonium and of highly enriched uranium

    International Nuclear Information System (INIS)

    The Director General has received a Note Verbale, dated 17 July 2003, from the Permanent Mission of the United Kingdom of Great Britain and Northern Ireland to the IAEA in the enclosures of which the Government of the United Kingdom, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for its national holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2002. The Government of the United Kingdom has also made available a statement of its annual figures for holdings of civil high enriched uranium (HEU), and of civil depleted, natural and low enriched uranium (DNLEU) in the civil nuclear fuel cycle, as of 31 December 2002. 3. In the light of the requests expressed by the Government of the United Kingdom in its Note Verbale of 1 December 1997 concerning its policies regarding the management of plutonium (INFCIRC/549 of 16 March 1998) and in its Note Verbale of 17 July 2003, the Note Verbale of 17 July 2003 and the enclosures thereto are attached for the information of all Member States

  19. Communication Received from the United Kingdom of Great Britain and Northern Ireland Concerning its Policies Regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Secretariat has received a note verbale dated 6 June 2011 from the Permanent Mission of the United Kingdom of Great Britain and Northern Ireland to the IAEA in the enclosures of which the Government, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines') and in accordance with Annexes B and C of the Guidelines, has made available annual figures for its holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2010. The Government of the United Kingdom has also made available a statement of its annual figures for holdings of civil high enriched uranium and of civil depleted, natural and low enriched uranium in the civil nuclear fuel cycle as of 31 December 2010. In light of the requests expressed by the Government of the United Kingdom in its note verbale of 1 December 1997 concerning its policies regarding the management of plutonium (INFCIRC/549 of 16 March 1998) and in its note verbale of 6 June 2011, the note verbale and the enclosures thereto are attached for the information of all Member States

  20. Formation of plutonium phosphates in chloride melts

    Energy Technology Data Exchange (ETDEWEB)

    Burnaeva, A.A.; Kryukova, A.I.; Kazanstev, G' N.; Skiba, O.V.; Korshunov, I.A.

    1984-01-01

    Introduction of sodium- and potassium phosphates Na/sub 3/PO/sub 4/ and K/sub 3/PO/sub 4/ in the PuCl/sub 3/-NaCl, PuCl/sub 3/-KCl melts results in reduction of plutonium amount in the liquid phase. Low-soluble plutonium (3) phosphates, of assumed Na/sub 3/Pu/sub 2/ composition (PO/sub 4/)/sub 3/ are transported into the solid phase. Using the methods of radiographical and radiometric analyses the phases of plutonium phosphates separated by precipitation from chloride melt and also prepared from PuO/sub 2/ and NaH/sub 2/PO/sub 4/ at 1200 deg C are investigated. Their solubility in the NaCl-KCl melt and stability to these melts during a long-term contact, and also under the effect of CCl/sub 4/ are evaluated. The data are compared with similar data for thorium-, uranium-, americium-, curium-, zirconium-, rare earth phosphates.

  1. Measurement of trace uranium-235 and plutonium-239, 240 in waste tank material at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Mahannah, F.N.; Maxwell, S.L. III.

    1992-01-01

    At the Savannah River Site (SRS), large quantities of radioactive liquid waste are evaporated to reduce volume before eventual processing through the In-Tank Precipitation process. Actinides in the liquid waste are only slightly soluble in the highly alkaline waste solution. Since some of the actinide isotopes are fissionable, the quantities being processed through the evaporator system are of interest. To better quantify the concentration and mass of fissionable material entering the evaporator system and eventually deposited as salt, analysis of the actinide elements were necessary. The predominant fissionable actinide isotopes of interest are U{sup 235} and Pu{sup 239}. To enable the reliable measurement of these radionuclides, the Central Laboratory has developed high speed separation techniques to measure U{sup 235} content by Isotope Dilution Mass Spectrometry and Pu{sup 239,240} by alpha spectrometry. Due to the high radioactivity levels in the samples all separations are performed in shielded analytical cells. Uranium is purified and concentrated using a high speed extraction chromatography technique that employs applied vacuum and columns containing tri (2-ethylene) phosphate solvent coated on a small particle inert support. The uranium method enables measurement of U{sup 235} concentrations to 1 {times} 10{sup {minus}4} g/L. Plutonium is purified and concentrated using a high speed anion exchange technique. The Pu method enables measurements of Pu{sup 239,240} to 2 {times} 10{sup {minus}6} g/L.

  2. Measurement of trace uranium-235 and plutonium-239, 240 in waste tank material at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Mahannah, F.N.; Maxwell, S.L. III

    1992-08-01

    At the Savannah River Site (SRS), large quantities of radioactive liquid waste are evaporated to reduce volume before eventual processing through the In-Tank Precipitation process. Actinides in the liquid waste are only slightly soluble in the highly alkaline waste solution. Since some of the actinide isotopes are fissionable, the quantities being processed through the evaporator system are of interest. To better quantify the concentration and mass of fissionable material entering the evaporator system and eventually deposited as salt, analysis of the actinide elements were necessary. The predominant fissionable actinide isotopes of interest are U{sup 235} and Pu{sup 239}. To enable the reliable measurement of these radionuclides, the Central Laboratory has developed high speed separation techniques to measure U{sup 235} content by Isotope Dilution Mass Spectrometry and Pu{sup 239,240} by alpha spectrometry. Due to the high radioactivity levels in the samples all separations are performed in shielded analytical cells. Uranium is purified and concentrated using a high speed extraction chromatography technique that employs applied vacuum and columns containing tri (2-ethylene) phosphate solvent coated on a small particle inert support. The uranium method enables measurement of U{sup 235} concentrations to 1 {times} 10{sup {minus}4} g/L. Plutonium is purified and concentrated using a high speed anion exchange technique. The Pu method enables measurements of Pu{sup 239,240} to 2 {times} 10{sup {minus}6} g/L.

  3. Measurement of trace uranium-235 and plutonium-239, 240 in waste tank material at the Savannah River Site

    International Nuclear Information System (INIS)

    At the Savannah River Site (SRS), large quantities of radioactive liquid waste are evaporated to reduce volume before eventual processing through the In-Tank Precipitation process. Actinides in the liquid waste are only slightly soluble in the highly alkaline waste solution. Since some of the actinide isotopes are fissionable, the quantities being processed through the evaporator system are of interest. To better quantify the concentration and mass of fissionable material entering the evaporator system and eventually deposited as salt, analysis of the actinide elements were necessary. The predominant fissionable actinide isotopes of interest are U235 and Pu239. To enable the reliable measurement of these radionuclides, the Central Laboratory has developed high speed separation techniques to measure U235 content by Isotope Dilution Mass Spectrometry and Pu239,240 by alpha spectrometry. Due to the high radioactivity levels in the samples all separations are performed in shielded analytical cells. Uranium is purified and concentrated using a high speed extraction chromatography technique that employs applied vacuum and columns containing tri (2-ethylene) phosphate solvent coated on a small particle inert support. The uranium method enables measurement of U235 concentrations to 1 x 10-4 g/L. Plutonium is purified and concentrated using a high speed anion exchange technique. The Pu method enables measurements of Pu239,240 to 2 x 10-6 g/L

  4. Plutonium in plants

    International Nuclear Information System (INIS)

    A bibliography on plutonium in plants is presented. It covers the subjects occurrence of plutonium in plants; soil-plant relationships; root uptake; distribution and translocation; foliar deposition and loss. Compiled data are presented on: recorded and calculated concentration factors of plutonium as well as those for uranium; concentration ratios for several crop types; proportion of plutonium removed from soil by plants; concentration ratios according to plant parts of cereal and vegetable crops. (G.J.P.)

  5. Estimated inventory of plutonium and uranium radionuclides for vegetation in aged fallout areas

    International Nuclear Information System (INIS)

    Data are presented pertinent to the contamination of vegetation by plutonium and other radionuclides in aged fallout areas on the Nevada Test Site (NTS) and the Tonopah Test Range (TTR). The standing biomass of vegetation estimated by nondestructive dimensional methods varied from about 200 to 600 g/m2 for the different fallout areas. Estimated inventories of 238Pu, 239Pu, 240Pu, and 235U in plants and their biological effects are discussed

  6. Preparation of americium amalgam

    International Nuclear Information System (INIS)

    Using the method of NGR-spectroscopy with the aid of 241Am isotope chemical state of transuranium elements in the volume and on the surface of amalgams is studied. Amalgam preparation was realized in a simplified electrolytic cell. It is shown that in the process of amalgam preparation the first order of reaction as to actinide is observed; americium is distributed gradually over the volume and it is partially sorbed by the surface of glass capillary. NGR spectrum of dry residue after mercury distillation at 200 deg C points to the presence of americium-mercury intermetal compounds

  7. Direct isotope ratio analysis of individual uranium–plutonium mixed particles with various U/Pu ratios by thermal ionization mass spectrometry

    International Nuclear Information System (INIS)

    Uranium and plutonium isotope ratios in individual uranium–plutonium (U–Pu) mixed particles with various U/Pu atomic ratios were analyzed without prior chemical separation by thermal ionization mass spectrometry (TIMS). Prior to measurement, micron-sized particles with U/Pu ratios of 1, 5, 10, 18, and 70 were produced from uranium and plutonium certified reference materials. In the TIMS analysis, the peaks of americium, plutonium, and uranium ion signals were successfully separated by continuously increasing the evaporation filament current. Consequently, the uranium and plutonium isotope ratios, except the 238Pu/239Pu ratio, were successfully determined for the particles at all U/Pu ratios. This indicates that TIMS direct analysis allows for the measurement of individual U–Pu mixed particles without prior chemical separation. - Highlights: • U–Pu mixed particles with U/Pu ratios of 1, 5, 10, 18 and 70 were produced. • U and Pu isotope ratios were determined by TIMS for particles without prior chemical separation. • Accurate values were obtained for all isotope ratios, except the 238Pu/239Pu ratio

  8. Optimisation and application of ICP-MS and alpha-spectrometry for determination of isotopic ratios of depleted uranium and plutonium in samples collected in Kosovo

    OpenAIRE

    Boulyga, S. F.; Testa, C; Desideri, D.; Becker, J. S.

    2001-01-01

    The determination of environmental contamination with natural and artificial actinide isotopes and evaluation of their source requires precise isotopic determination of actinides, above all uranium and plutonium. This can be achieved by alpha spectrometry or by inductively coupled plasma mass spectrometry (ICP-MS) after chemical separation of actinides. The performance of a sector-field ICP-MS (ICP-SFMS) coupled to a low-flow micronebulizer with a membrane desolvation unit, "Aridus'', was stu...

  9. Process for separation of technetium from zirconium in an organic solvent and at least another metal such as uranium or plutonium for reprocessing spent nuclear fuels

    International Nuclear Information System (INIS)

    In an organic solution containing U, Pu, Zr and Tc, zirconium is extracted with a solution of nitric acid and/or nitrates, then technetium is extracted with a solution of nitric acid and/or nitrates at a different concentration. In this way Tc is extracted in the initial step of reprocessing before uranium, plutonium separation without special chemicals and without increasing too much aqueous effuents

  10. In Plant Measurement and Analysis of Mixtures of Uranium and Plutonium TRU-Waste Using a {sup 252}Cf Shuffler Instrument

    Energy Technology Data Exchange (ETDEWEB)

    Hurd, J.R.

    1998-11-02

    The active-passive {sup 252}Cf shuffler instrument, installed and certified several years ago in Los Alamos National Laboratory's plutonium facility, has now been calibrated for different matrices to measure Waste Isolation Pilot Plant (WIPP)-destined transuranic (TRU)-waste. Little or no data currently exist for these types of measurements in plant environments where sudden large changes in the neutron background radiation can significantly distort the results. Measurements and analyses of twenty-two 55-gallon drums, consisting of mixtures of varying quantities of uranium and plutonium in mostly noncombustible matrices, have been recently completed at the plutonium facility. The calibration and measurement techniques, including the method used to separate out the plutonium component, will be presented and discussed. Calculations used to adjust for differences in uranium enrichment from that of the calibration standards will be shown. Methods used to determine various sources of both random and systematic error will be indicated. Particular attention will be directed to those problems identified as arising from the plant environment. The results of studies to quantify the aforementioned distortion effects in the data will be presented. Various solution scenarios will be outlined, along with those adopted here.

  11. Chemistry of americium

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, W.W.

    1976-01-01

    Essential features of the descriptive chemistry of americium are reviewed. Chapter titles are: discovery, atomic and nuclear properties, collateral reading, production and uses, chemistry in aqueous solution, metal, alloys, and compounds, and, recovery, separation, purification. Author and subject indexes are included. (JCB)

  12. Uranium and plutonium in marine sediments; Uranio y plutonio en sedimentos marinos

    Energy Technology Data Exchange (ETDEWEB)

    Ordonez R, E.; Almazan T, M. G. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ruiz F, A. C., E-mail: eduardo.ordonez@inin.gob.mx [UNAM, Instituto de Ciencias del Mar y Limnologia, Unidad Academica Mazatlan, Sinaloa (MX)

    2011-11-15

    The marine sediments contain uranium concentrations that are considered normal, since the seawater contains dissolved natural uranium that is deposited in the bed sea in form of sediments by physical-chemistry and bio-genetics processes. Since the natural uranium is constituted of several isotopes, the analysis of the isotopic relationship {sup 234}U/{sup 238}U are an indicator of the oceanic activity that goes accumulating slowly leaving a historical registration of the marine events through the profile of the marine soil. But the uranium is not the only radioelement present in the marine sediments. In the most superficial strata the presence of the {sup 239+140}Pu has been detected that it is an alpha emitter and that recently it has been detected with more frequency in some coasts of the world. The Mexican coast has not been the exception to this phenomenon and in this work the presence of {sup 239-140}Pu is shown in the more superficial layers of an exploring coming from the Gulf of Tehuantepec. (Author)

  13. Non-destructive assay system for uranium and plutonium in reprocessing input solutions. Hybrid K-edge/XRF Densitometer. JASPAS JC-11 final report

    International Nuclear Information System (INIS)

    As a part of JASPAS programme, a non-radioactive assay system for the accountability of uranium and plutonium in input dissolver solutions of a spent fuel reprocessing plant, called Hybrid K-edge/XRF Densitometer, has been developed at the Tokai Reprocessing plant (TRP) since 1991. The instrument is the one of the hybrid type combined K-edge densitometry (KED) and X-ray fluorescence (XRF) analysis. The KED is used to determine the uranium concentration and the XRF is used to determine the U/Pu ratio. These results give the plutonium concentration in consequence. It is considered that the instrument has the capability of timely on-site verification for input accountancy. The instrument had been installed in the analytical hot cell at the TRP and the experiments comparing with Isotope Dilution Mass Spectrometry (IDMS) method have been carried out. As the results of measurements for the actual input solutions in the acceptance and performance tests, it was typically confirmed that the precision for determining uranium concentration by the KED was within 0.2%, whereas the XRF for plutonium performed within 0.7%. This final report summarizes the design information and performance data so as to end the JASPAS programme. (author)

  14. Non-destructive assay system for uranium and plutonium in reprocessing input solutions. Hybrid K-edge/XRF Densitometer. JASPAS JC-11 final report

    Energy Technology Data Exchange (ETDEWEB)

    Surugaya, N.; Abe, K.; Kurosawa, A.; Ikeda, H.; Kuno, Y.

    1997-05-01

    As a part of JASPAS programme, a non-radioactive assay system for the accountability of uranium and plutonium in input dissolver solutions of a spent fuel reprocessing plant, called Hybrid K-edge/XRF Densitometer, has been developed at the Tokai Reprocessing plant (TRP) since 1991. The instrument is the one of the hybrid type combined K-edge densitometry (KED) and X-ray fluorescence (XRF) analysis. The KED is used to determine the uranium concentration and the XRF is used to determine the U/Pu ratio. These results give the plutonium concentration in consequence. It is considered that the instrument has the capability of timely on-site verification for input accountancy. The instrument had been installed in the analytical hot cell at the TRP and the experiments comparing with Isotope Dilution Mass Spectrometry (IDMS) method have been carried out. As the results of measurements for the actual input solutions in the acceptance and performance tests, it was typically confirmed that the precision for determining uranium concentration by the KED was within 0.2%, whereas the XRF for plutonium performed within 0.7%. This final report summarizes the design information and performance data so as to end the JASPAS programme. (author)

  15. Evaluation of Software to Nondestructively Quantify Uranium and Plutonium Samples via Gamma Ray Spectroscopy at the Joint Research Center in Ispra, Italy

    International Nuclear Information System (INIS)

    In March, 2001 at the Joint Research Center in Ispra Italy, an evaluation measurement exercise was undertaken by the Inst. de Radioprotection et de Surete nucleaire (IRSN), for the evaluation of software used for the nondestructive analysis of special nuclear material held by the Joint Research Center. Staff of IRSN and ORTEC measured and analyzed well-characterized uranium and plutonium samples at the JRC PERLA facility, Ispra. Three of the codes evaluated were GammaVision (Version 5.1), 1 ISOTOPIC (Version 2.0.6.2), 2 and PC/FRAM (Version 2.3). 3 GammaVision was used to determine the primary activities relative to a certified calibration point source. ISOTOPIC used modeling to compute correction factors for attenuation and geometry to determine gram quantities of each nuclide following the GammaVision analysis. PC/FRAM was used to determine isotopic ratios or weight percent of the plutonium isotopes. Information from this exercise was used to determine accuracy estimates for inspection teams who would use these methods to measure and analyze uranium and plutonium samples of similar composition at French nuclear sites. Samples selected by the Joint Research Center were measured and analyzed as unknowns. Average uncertainties were determined for homogenous material consisting of UO2, U3O8, PuO2 and MOX (mixed oxides of uranium and plutonium). Several nonhomogeneous samples were prepared in order to simulate containers of radioactive waste. The GammaVision measurement system was calibrated using a traceable 152Eu point source. Geometry corrections were then performed to relate the item being measured to the point source. Isotopic used these results and modeled the actual sample to realistically reflect the volume and attenuation characteristics. ISOTOPIC possesses a 'fine-tune' adjustment: which assumes that activity calculated from individual lines of nuclides with multiple gamma rays must be the same if the correction factors for density, weight fraction uranium

  16. Scope of STACY experiments for criticality benchmark data on low enriched uranium and plutonium solution system

    International Nuclear Information System (INIS)

    The Static Experiment Critical Facility, STACY was constructed in the Nuclear Fuel Cycle Safety Engineering Research Facility, NUCEF of the Japan Atomic Energy Research Institute in order to produce the fundamental critical data of uranyl nitrate solution, plutonium nitrate solution and their mixture. A series of experiments using single core tank have been performed using 10% enriched uranyl nitrate solution since the first criticality in 1995. Benchmark data of STACY are now used for verification of Japanese criticality safety code system and nuclear data libraries. Kinetic parameters, temperature coefficients and reflector effects of structural material are also measured using single homogeneous core. It is on schedule to make experiments for neutron interaction effect and for simulating the dissolving process with a heterogeneous core using low enriched uranyl nitrate solution. After these experiments, systematic critical and subcritical experiments on plutonium nitrate solution will start in five years. This paper reviews the main results of STACY since the initial criticality and describes the criticality properties of the experimental cores in the future program. (author)

  17. Communication received from the Government of the Federal Republic of Germany concerning its policies regarding the management of plutonium. Statements on the management of plutonium and of high enriched uranium

    International Nuclear Information System (INIS)

    The Director General has received a Note Verbale, dated 17 September 2004, from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2003. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of highly enriched uranium (HEU) as of 31 December 2003. In light of the request expressed by the Federal Republic of Germany in its Note Verbale of 1 December 1997 concerning its policies regarding the management of plutonium (INFCIRC/549 of 16 March 1998), the Note Verbale of 17 September 2004 and the enclosures thereto are attached for the information of all Member States

  18. Communication received from the Government of the Federal Republic of Germany concerning its policies regarding the management of plutonium. Statements on the management of plutonium and of high enriched uranium

    International Nuclear Information System (INIS)

    The Director General has received a Note Verbale, dated 22 September 2003, from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2002. 2. The Government of the Federal Republic of Germany has also made available a statement of the estimated amounts of high enriched uranium (HEU) as of 31 December 2002. 3. In light of the request expressed by the Federal Republic of Germany in its Note Verbale of 1 December 1997 concerning its policies regarding the management of plutonium (INFCIRC/549 of 16 March 1998), the Note Verbale of 22 September 2003 and the enclosures thereto are attached for the information of all Member States

  19. Extraction and separation of uranium, plutonium and americium from MOX fuel rejected waste and analytical phosphoric acid based waste using organophosphorus extractants

    International Nuclear Information System (INIS)

    The extraction of U, Pu and Am from HNO3 solutions using tri-n-butyl phosphate (TBP), a mixture of 0.2M CMPO + 1.2M TBP and Cyanex-923 in dodecane has been carried out. The mixed oxide (MOX) fuel rejected during fabrication, termed as dirty rejected oxide (DRO) sludge was dissolved in HNO3 and a procedure has been developed to recover U, Pu and Am from this waste using the above extractants. Also, a procedure has been developed to recover almost quantitatively the U, Pu and Am content from one of the difficult analytical wastes generated after the Davies and Gray analysis of U, in presence of Pu containing H3PO4, H2SO4, and HNO3, by using a mixture of 30% Cyanex-923 + 20% TBP in dodecane. (author)

  20. The uranium-carbon and plutonium-carbon systems. A thermochemical assessment

    International Nuclear Information System (INIS)

    A fair amount of thermochemical data has been accumulated on the compounds in the uranium-carbon system. The main difficulties involved appear to be the sluggishness of the reaction of these carbides and the lack of information on the true equilibrium diagram. The information assessed in this report is accurate to, say ± 5 kcal on the average. This is in fact satisfactory for quite a number of calculations of equilibria involving uranium and carbon. It is not accurate enough for more ambitious calculations such as that of the equilibrium diagram. Present assessment has also made clear the gaps that still exist. It appears that it is mainly the non-stoichiometric parts of the diagram that need extensive further studies; this would also assist in increasing the accuracy of the known data. 66 refs, 6 figs, 15 tabs

  1. Uranium, plutonium, and thorium isotopes in the atmosphere and the lithosphere

    International Nuclear Information System (INIS)

    Concentration of 238U in rain and snow collected at Fayetteville (360N, 940W), Arkansas, showed a marked increase during the summer months of 1980, while Mount St. Helens remained active. This observed increase of 238U can be explained as due to the fallout of natural uranium from the eruption of Mount St. Helens. Large increases in the concentration of thorium isotopes detected in rain and snow samples during the last months of 1982 and early months of 1983 probably originated from the eruption of El Chichon volcano, which occurred on 28 March 1982. About 450 Ci of 232Th is estimated to have been injected into the atmosphere by this eruption. Isotopic anomalies were observed in atmospheric samples such as rain and snow. These anomalies can be attributed to various natural as well as man-made sources: nuclear weapon tests, nuclear accidents involving the burn-up of nuclear powered satellites, and volcanic eruptions. The variation of 234U/238U ratios in radioactive minerals when leached with nitric acid were also noticed and this variation, while 235U/238U remained fairly constant, can be explained in terms of the α-recoil effect and changes in oxidation state of uranium. Difference found in 239Pu/238U ratios in terrestrial samples and uranium minerals can be explained as due to fallout contamination

  2. Plutonium-236 traces determination in plutonium-238 by α spectrometry

    International Nuclear Information System (INIS)

    Two methods are described in this report for the determination of plutonium-236 traces in plutonium-238 by a spectrometry using semi-conductor detectors. The first method involves a direct comparison of the areas under the peaks of the α spectra of plutonium-236 and plutonium-238. The electrolytic preparation of the sources is carried out after preliminary purification of the plutonium. The second method makes it possible to determine the 236Pu/238Pu ratio by comparing the areas of the α peaks of uranium-232 and uranium-234, which are the decay products of the two plutonium isotopes respectively. The uranium in the source, also deposited by electrolysis, is separated from a 1 mg amount of plutonium either by a T.L.A. extraction, or by the use of ion-exchange resins. The report ends with a discussion of the results obtained with plutonium of two different origins. (authors)

  3. Communication Received from Germany Concerning its Policies Regarding the Management of Plutonium. Statements on the Management of Plutonium and of Highly Enriched Uranium

    International Nuclear Information System (INIS)

    The Secretariat has received a letter dated 23 June 2006 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2005

  4. Communication Received from Germany Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of Highly Enriched Uranium

    International Nuclear Information System (INIS)

    The Director General has received a letter dated 20 June 2008 from the Permanent Mission of the Federal Republic of Germany to the IAEA in the enclosures of which the Government of Germany, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2007

  5. Communication Received from France Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Director General has received a note verbale dated 14 June 2012 from the Permanent Mission of France to the IAEA in the enclosures of which the Government of France, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2011

  6. Communication Received from France Concerning Its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Secretariat has received a note verbale dated 2 May 2013 from the Permanent Mission of Switzerland to the IAEA in the enclosures of which the Government of Switzerland, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as 'Guidelines') and in accordance with Annexes B and C of the Guidelines, has made available annual figures for its holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2012

  7. Communication Received from France Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Director General has received a note verbale dated 31 July 2013 from the Permanent Mission of France to the IAEA in the enclosures of which the Government of France, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2012

  8. Communication Received from France Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Director General has received a note verbale dated 16 September 2011 from the Permanent Mission of France to the IAEA in the enclosures of which the Government of France, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/5491 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2010

  9. Communication Received from France Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Director General has received a note verbale dated 24 August 2010 from the Permanent Mission of France to the IAEA in the enclosures of which the Government of France, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2009

  10. Communication Received from France Concerning its Policies regarding the Management of Plutonium. Statements on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Director General has received a Note Verbale dated 17 July 2007 from the Permanent Mission of France to the IAEA in the enclosures of which the Government of France, in keeping with its commitment under the Guidelines for the Management of Plutonium (contained in INFCIRC/549 of 16 March 1998 and hereinafter referred to as the 'Guidelines'), and in accordance with Annexes B and C of the Guidelines, has made available annual figures for holdings of civil unirradiated plutonium and the estimated amounts of plutonium contained in spent civil reactor fuel as of 31 December 2006

  11. Mound Laboratory activities on the removal of plutonium and uranium from wastewater using bone char

    Energy Technology Data Exchange (ETDEWEB)

    Blane, D.W.; Murphy, E.L.

    1976-09-30

    Pilot plant studies sponsored by the Division of Military Application (DMA) to treat Mound Laboratory's low risk waste streams with bone char columns have been completed. This project was to accomplish the following: (1) acquire engineering data such as flow rates, adsorption capacity, and optimum particle size for designing a tertiary treatment system, and (2) optimize the primary waste treatment process using coagulants and flocculants. Bone char, a natural product made from granulated cattle bone, is a form of calcium hydroxyapatite (Ca/sub 10/(PO/sub 4/)/sub 6/(OH)/sub 2/) and contains a small amount of carbon. It is characterized by a high porosity and good resistance to abrasion and crushing. Because plutonium and other actinides can be precipitated as phosphates from weakly acidic, neutral or alkaline solutions, it appeared possible to use an insoluble phosphate adsorbent such as bone char to remove them from waste streams.

  12. Experience of determination of plutonium and uranium contents in MOX fuel by IDMS

    International Nuclear Information System (INIS)

    In the Plutonium Fuel Center (PFC) of JNC, Isotope Dilution Mass Spectrometry (IDMS) has been used to determine Pu and U contents of nuclear materials since 1996. In MOX fabrication plant, many types of sample with wide variation of Pu/U ratio including aged Pu and process scrap should be analyzed for not only quality control purpose but also material accountancy. Because IDMS can eliminate influences of coexistence elements and has high accuracy, it is considered to be the best analytical method for MOX fabrication plant. This paper summarizes the experience of IDMS in the PFC laboratory including the preparation of Large Size Dried (LSD) spike, and also describes the evaluation of analytical error and consideration on procurement of LSD spike for IDMS

  13. Separation of plutonium from uranium and fission products in the zirconium pyrophospate column

    International Nuclear Information System (INIS)

    Distribution coefficients were of the following ions were determined in the system zirconium pyrophosphate - aqueous solution HNO3 : Pu3+, Pu4+, PuO22+, UO22+, 234Th2+, 95Zr, 95Nb, 106Ru, 144Ce3+, 90Sr2+, 137Cs+, 59Fe3+ and 59Fe2+. According to the distribution coefficients it can be concluded that the separation of some cations is possible. This was proved by using separation columns. The following successful separations were completed: 90Sr2+ from 90I3+, 90Sr2+ from 90I3+ and 1'37Cs+, UO2+ from 234Th4+, Pu4+ from UO22+, 95Zr, 95Nb, 106Ru, 144Ce3+, 90Sr2+, 137Cs+. Decontamination factors of plutonium from the mentioned cations were determined. It was found that the sorption of Cs+ and Sr2+ is based on ion exchange

  14. Evaluation of the characteristics of uranium and plutonium Mixed Oxide (MOX) fuel

    International Nuclear Information System (INIS)

    MOX fuel irradiation test up to high burnup has been performed for five years. Irradiation test of MOX fuel having high plutonium content has also been performed from JFY 2007 and it still continues. A lot of irradiation data have been obtained through these tests. The activities done in JFY 2012 are mainly focused on Post Irradiation Examination (PIE) data analysis concerning thermal property change and fission gas release. In the former work thermal conductivity degradation due to burnup is examined and in the latter work the dependence of fission gas release mechanism on fuel pellet microstructure is examined. This report mainly covers the result of analysis. It is found that thermal conductivity degradation of MOX fuel due to burnup is less than that of UO2 fuel and that fission gas release mechanism of high enriched fissile zone (so called Pu spot) is much different from that of low enriched fissile zone (so called Matrix). (author)

  15. Post irradiation examinations of uranium-plutonium mixed nitride fuel irradiated in JMTR (88F-5A capsule)

    International Nuclear Information System (INIS)

    Two helium-bonded fuel pins filled with uranium-plutonium mixed nitride pellets were encapsulated in 88F-5A and irradiated in JMTR up to 4.1%FIMA at a maximum linear power of 65 kW/m. The capsule cooled for ∼4 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pins. It was found from the temperature change at pellet center that the helium gap between the pellets and the cladding tube was gradually closed during irradiation. Very low fission gas release rate of about 2-3% was observed, while the diametric increase of fuel pin was limited to ∼0.4% at the position of maximum reading. The inner surface of cladding tube did not show the signs of chemical interaction with fuel. In addition, some information on restructuring of fuel pellets was obtained. (author)

  16. Nuclear weapon relevant materials and preventive arms control. Uranium-free fuels for plutonium elimination and spallation neutron sources

    International Nuclear Information System (INIS)

    Today, the most significant barrier against the access to nuclear weapons is to take hold on sufficient amounts of nuclear weapon-relevant nuclear materials. It is mainly a matter of fissionable materials (like highly enriched uranium and plutonium) but also of fusionable tritium. These can be used as reactor fuel in civil nuclear programmes but also in nuclear weapon programmes. To stop or to hinder nuclear proliferation, in consequence, there is not only a need to analyse open or covered political objectives and intentions. In the long term, it might be more decisive to analyse the intrinsic civil-military ambivalence of nuclear materials and technologies, which are suitable for sensitive material production. A farsighted strategy to avoid proliferation dangers should take much more account to technical capabilities as it is done in the political debate on nuclear non-proliferation so far. If a technical option is at a state's disposal, it is extremely difficult and lengthy to revert that again. The dangers, which one has to react to, are stemming from already existing stocks of nuclear weapon-relevant materials - in the military as well as in the civil realm - and from existing or future technologies, which are suitable for the production of such materials (cf. info 1 and 2). Therefore, the overall approach of this research project is to strive for a drastic reduction of the access to nuclear weapon-relevant material and its production capabilities. Thus, on one hand the nuclear proliferation by state actors could be answered more effectively, on the other hand by that approach a decisive barrier against the access on nuclear weapons by sub-national groups and terrorists could also be erected. For this purpose, safeguards of the International Atomic Energy Agency (IAEA) and other measures of physical accountancy will remain indispensable elements of arms control. However, one has to consider that the goal of nuclear non-proliferation could not be achieved and

  17. Evaluation of the Integrated Holdup Measurement System with the M3CA for Assay of Uranium and Plutonium Holdup

    International Nuclear Information System (INIS)

    Uranium and plutonium holdup that has been simulated by insertion of a variety of sealed, reference samples into pipes, ducts, and other hardware has been measured over a period of six years with an integrated holdup measurement system. The result is a systematic evaluation of the generalized-geometry holdup (GGH) formalism applied to portable gamma-ray holdup measurements with low-resolution detectors. The extended exercise was carried out both with and without automation of the measurements, data reduction/analysis, and holdup evaluation. Automation was accomplished by the software Version 2 for the Holdup Measurement System (HMS2). The purpose of the exercise was to establish reliable benchmarks for GGH measurements and to document the advantages of the automation with actual measurement results. The results presented below demonstrate a factor of 2 improvement in the quantitative reliability of the holdup assay automated by HMS2. The automated results are otherwise identical to the manual measurements. These and similar exercises also show that automation can decrease by a factor of 20 or more the time required to execute a holdup measurement campaign and obtain the holdup quantities for the facility using an integrated holdup measurement system, and that only one person, rather than two, is required to perform the measurements. Enhanced implementation of the integrated holdup measurement system with new software, corrections for systematic effects, and improved room-temperature gamma-ray detectors is planned

  18. Potential for radionuclide immobilization in the EBS/NFE: solubility limiting phases for neptunium, plutonium, and uranium

    Energy Technology Data Exchange (ETDEWEB)

    Rard, J. A., LLNL

    1997-10-01

    Retardation and dispersion in the far field of radionuclides released from the engineered barrier system/near field environment (EBS/NFE) may not be sufficient to prevent regulatory limits being exceeded at the accessible environment. Hence, a greater emphasis must be placed on retardation and/or immobilization of radionuclides in the EBS/NFE. The present document represents a survey of radionuclide-bearing solid phases that could potentially form in the EBS/NFE and immobilize radionuclides released from the waste package and significantly reduce the source term. A detailed literature search was undertaken for experimental solubilities of the oxides, hydroxides, and various salts of neptunium, plutonium, and uranium in aqueous solutions as functions of pH, temperature, and the concentrations of added electrolytes. Numerous solubility studies and reviews were identified and copies of most of the articles were acquired. However, this project was only two months in duration, and copies of some the identified solubility studies could not be obtained at short notice. The results of this survey are intended to be used to assess whether a more detailed study of identified low- solubility phase(s) is warranted, and not as a data base suitable for predicting radionuclide solubility. The results of this survey may also prove useful in a preliminary evaluation of the efficacy of incorporating chemical additives to the EBS/NFE that will enhance radionuclide immobilization.

  19. Controlling the oxygen potential to improve the densification and the solid solution formation of uranium-plutonium mixed oxides

    Science.gov (United States)

    Berzati, Ségolène; Vaudez, Stéphane; Belin, Renaud C.; Léchelle, Jacques; Marc, Yves; Richaud, Jean-Christophe; Heintz, Jean-Marc

    2014-04-01

    Diffusion mechanisms occurring during the sintering of oxide ceramics are affected by the oxygen content of the atmosphere, as it imposes the nature and the concentration of structural defects in the material. Thus, the oxygen partial pressure, p(O2), of the sintering gas has to be precisely controlled, otherwise a large dispersion in various parameters, critical for the manufacturing of ceramics such as nuclear oxides fuels, is likely to occur. In the present work, the densification behaviour and the solid solution formation of a mixed uranium-plutonium oxide (MOX) were investigated. The initial mixture, composed of 70% UO2 + 30% PuO2, was studied at p(O2) ranging from 10-15 to 10-4 atm up to 1873 K both with dilatometry and in situ high temperature X-ray diffraction. This study has shown that the initial oxides UO2+x and PuO2-x first densify during heating and then the solid solution formation starts at about 200 K higher. The densification and the formation of the solid solution both occur at a lower temperature when p(O2) increases. Based on this result, it is possible to better define the sintering atmosphere, eventually leading to optimized parameters such as density, oxygen stoichiometry and cations homogenization of nuclear ceramics and of a wide range of industrial ceramic materials.

  20. Uranium and Plutonium Average Prompt-fission Neutron Energy Spectra (PFNS) from the Analysis of NTS NUEX Data

    Science.gov (United States)

    Lestone, J. P.; Shores, E. F.

    2014-05-01

    In neutron experiments (NUEX) conducted at the Nevada Test Site (NTS) by Los Alamos National Laboratory, the time-of-flight of fission-neutrons emitted from nuclear tests were observed by measuring the current generated by the collection of protons scattered from a thin CH2 foil many meters from the nuclear device into a Faraday cup. The time dependence of the Faraday cup current is a measure of the energy spectrum of the neutrons that leak from the device. With good device models and accurate neutron-transport codes, the leakage spectra can be converted into prompt fast-neutron-induced fission-neutron energy spectra. This has been done for two events containing plutonium, and for an earlier event containing uranium. The prompt-fission neutron spectra have been inferred for 1.5-MeV 239Pu(n,f) and 235U(n,f) reactions for outgoing neutron energies from 1.5 to ∼10.5 MeV, in 1-MeV steps. These spectra are in good agreement with the Los Alamos fission model.