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Sample records for americium plutonium uranium

  1. 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)

  2. 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

  3. LIBS Spectral Data for a Mixed Actinide Fuel Pellet Containing Uranium, Plutonium, Neptunium and Americium

    Energy Technology Data Exchange (ETDEWEB)

    Judge, Elizabeth J. [Los Alamos National Laboratory; Berg, John M. [Los Alamos National Laboratory; Le, Loan A. [Los Alamos National Laboratory; Lopez, Leon N. [Los Alamos National Laboratory; Barefield, James E. [Los Alamos National Laboratory

    2012-06-18

    Laser-induced breakdown spectroscopy (LIBS) was used to analyze a mixed actinide fuel pellet containing 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2}. The preliminary data shown here is the first report of LIBS analysis of a mixed actinide fuel pellet, to the authors knowledge. The LIBS spectral data was acquired in a plutonium facility at Los Alamos National Laboratory where the sample was contained within a glove box. The initial installation of the glove box was not intended for complete ultraviolet (UV), visible (VIS) and near infrared (NIR) transmission, therefore the LIBS spectrum is truncated in the UV and NIR regions due to the optical transmission of the window port and filters that were installed. The optical collection of the emission from the LIBS plasma will be optimized in the future. However, the preliminary LIBS data acquired is worth reporting due to the uniqueness of the sample and spectral data. The analysis of several actinides in the presence of each other is an important feature of this analysis since traditional methods must chemically separate uranium, plutonium, neptunium, and americium prior to analysis. Due to the historic nature of the sample fuel pellet analyzed, the provided sample composition of 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2} cannot be confirm without further analytical processing. Uranium, plutonium, and americium emission lines were abundant and easily assigned while neptunium was more difficult to identify. There may be several reasons for this observation, other than knowing the exact sample composition of the fuel pellet. First, the atomic emission wavelength resources for neptunium are limited and such techniques as hollow cathode discharge lamp have different dynamics than the plasma used in LIBS which results in different emission spectra. Secondly, due to the complex sample of four actinide elements, which all have very dense electronic energy levels, there may be reactions and

  4. 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)

  5. Functional sorbents for selective capture of plutonium, americium, uranium, and thorium in blood.

    Science.gov (United States)

    Yantasee, Wassana; Sangvanich, Thanapon; Creim, Jeffery A; Pattamakomsan, Kanda; Wiacek, Robert J; Fryxell, Glen E; Addleman, R Shane; Timchalk, Charles

    2010-09-01

    Self-assembled monolayer on mesoporous supports (SAMMS) are hybrid materials created from attachment of organic moieties onto very high surface area mesoporous silica. SAMMS with surface chemistries including three isomers of hydroxypyridinone, diphosphonic acid, acetamide phosphonic acid, glycinyl urea, and diethylenetriamine pentaacetate (DTPA) analog were evaluated for chelation of actinides ((239)Pu, (241)Am, uranium, thorium) from blood. Direct blood decorporation using sorbents does not have the toxicity or renal challenges associated with traditional chelation therapy and may have potential applications for critical exposure cases, reduction of nonspecific dose during actinide radiotherapy, and for sorbent hemoperfusion in renal insufficient patients, whose kidneys clear radionuclides at a very slow rate. Sorption affinity (K(d)), sorption rate, selectivity, and stability of SAMMS were measured in batch contact experiments. An isomer of hydroxypyridinone (3,4-HOPO) on SAMMS demonstrated the highest affinity for all four actinides from blood and plasma and greatly outperformed the DTPA analog on SAMMS and commercial resins. In batch contact, a fifty percent reduction of actinides in blood was achieved within minutes, and there was no evidence of protein fouling or material leaching in blood after 24 h. The engineered form of SAMMS (bead format) was further evaluated in a 100-fold scaled-down hemoperfusion device and showed no blood clotting after 2 h. A 0.2 g quantity of SAMMS could reduce 50 wt.% of 100 ppb uranium in 50 mL of plasma in 18 min and that of 500 dpm mL(-1) in 24 min. 3,4-HOPO-SAMMS has a long shelf-life in air and at room temperature for at least 8 y, indicating its feasibility for stockpiling in preparedness for an emergency. The excellent efficacy and stability of SAMMS materials in complex biological matrices suggest that SAMMS can also be used as orally administered drugs and for wound decontamination. By changing the organic groups of

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gardner, Kyle Shelton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); 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)

    2016-09-28

    These are a set of slides intended for an information session as part of recruiting activities at Brigham Young University. It gives an overview of aqueous chloride operations, specifically on plutonium and americium purification/recovery. This presentation details the steps taken perform these processes, from plutonium size reduction, dissolution, solvent extraction, oxalate precipitation, to calcination. For Americium recovery, it details the CLEAR (chloride extraction and actinide recovery) Line, oxalate precipitation and calcination.

  14. 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)

  15. 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)

  16. 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)

  17. 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.

  18. 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)

  19. 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

  20. Method for removal of plutonium impurity from americium oxides and fluorides

    Science.gov (United States)

    FitzPatrick, John R.; Dunn, Jerry G.; Avens, Larry R.

    1987-01-01

    Method for removal of plutonium impurity from americium oxides and fluorides. AmF.sub.4 is not further oxidized to AmF.sub.6 by the application of O.sub.2 F at room temperature, while plutonium compounds present in the americium sample are fluorinated to volatile PuF.sub.6, which can readily be separated therefrom, leaving the purified americium oxides and/or fluorides as the solid tetrafluoride.

  1. Uncertainty analysis of doses from ingestion of plutonium and americium.

    Science.gov (United States)

    Puncher, M; Harrison, J D

    2012-02-01

    Uncertainty analyses have been performed on the biokinetic model for americium currently used by the International Commission on Radiological Protection (ICRP), and the model for plutonium recently derived by Leggett, considering acute intakes by ingestion by adult members of the public. The analyses calculated distributions of doses per unit intake. Those parameters having the greatest impact on prospective doses were identified by sensitivity analysis; the most important were the fraction absorbed from the alimentary tract, f(1), and rates of uptake from blood to bone surfaces. Probability distributions were selected based on the observed distribution of plutonium and americium in human subjects where possible; the distributions for f(1) reflected uncertainty on the average value of this parameter for non-specified plutonium and americium compounds ingested by adult members of the public. The calculated distributions of effective doses for ingested (239)Pu and (241)Am were well described by log-normal distributions, with doses varying by around a factor of 3 above and below the central values; the distributions contain the current ICRP Publication 67 dose coefficients for ingestion of (239)Pu and (241)Am by adult members of the public. Uncertainty on f(1) values had the greatest impact on doses, particularly effective dose. It is concluded that: (1) more precise data on f(1) values would have a greater effect in reducing uncertainties on doses from ingested (239)Pu and (241)Am, than reducing uncertainty on other model parameter values and (2) the results support the dose coefficients (Sv Bq(-1) intake) derived by ICRP for ingestion of (239)Pu and (241)Am by adult members of the public.

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. Experimental studies and thermodynamic modelling of volatilities of uranium, plutonium, and americium from their oxides and from their oxides interacted with ash

    Energy Technology Data Exchange (ETDEWEB)

    Krikorian, O.H.; Ebbinghaus, B.B.; Adamson, M.G.; Fontes, A.S. Jr.; Fleming, D.L.

    1993-09-15

    The purpose of this study is to identify the types and amounts of volatile gaseous species of U, Pu, and Am that are produced in the combustion chamber offgases of mixed waste oxidation processors. Primary emphasis is on the Rocky Flats Plant Fluidized Bed Incinerator. Transpiration experiments have been carried out on U{sub 3}O{sub 8}(s), U{sub 3}O{sub 8} interacted with various ash materials, PuO{sub 2}(s), PuO{sub 2} interacted with ash materials, and a 3%PuO{sub 2}/0.06%AmO{sub 2}/ash material, all in the presence of steam and oxygen, and at temperatures in the vicinity of 1,300 K. UO{sub 3}(g) and UO{sub 2}(OH){sub 2}(g) have been identified as the uranium volatile species and thermodynamic data established for them. Pu and Am are found to have very low volatilities, and carryover of Pu and Am as fine dust particulates is found to dominate over vapor transport. The authors are able to set upper limits on Pu and Am volatilities. Very little lowering of U volatility is found for U{sub 3}O{sub 8} interacted with typical ashes. However, ashes high in Na{sub 2}O (6.4 wt %) or in CaO (25 wt %) showed about an order of magnitude reduction in U volatility. Thermodynamic modeling studies were carried out that show that for aluminosilicate ash materials, it is the presence of group I and group II oxides that reduces the activity of the actinide oxides. K{sub 2}O is the most effective followed by Na{sub 2}O and CaO for common ash constituents. A more major effect in actinide activity lowering could be achieved by adding excess group I or group II oxides to exceed their interaction with the ash and lead to direct formation of alkali or alkaline earth uranates, plutonates, and americates.

  7. 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.

  8. 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)

  9. 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.

  10. 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)

  11. 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)

  12. 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)

  13. 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

  14. 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...

  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. 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.

  17. 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.

  18. 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)

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

    Science.gov (United States)

    Van der Meeren, A; Grémy, O

    2010-09-01

    The biodistribution of plutonium and americium has been studied in a rat model after inhalation of two PuO(2) 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 PuO(2) with a higher proportion of americium shows an accelerated transfer of activity from lungs to blood as compared to PuO(2) 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 PuO(2) powders with a higher dissolution rate for both plutonium and americium being obtained for the PuO(2) 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 PuO(2) 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.

  20. 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

  1. 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.

  2. Calcium and zinc DTPA administration for internal contamination with plutonium-238 and americium-241.

    Science.gov (United States)

    Kazzi, Ziad N; Heyl, Alexander; Ruprecht, Johann

    2012-08-01

    The accidental or intentional release of plutonium or americium can cause acute and long term adverse health effects if they enter the human body by ingestion, inhalation, or injection. These effects can be prevented by rapid removal of these radionuclides by chelators such as calcium or zinc diethylenetriaminepentaacetate (calcium or zinc DTPA). These compounds have been shown to be efficacious in enhancing the elimination of members of the actinide family particularly plutonium and americium when administered intravenously or by nebulizer. The efficacy and adverse effects profile depend on several factors that include the route of internalization of the actinide, the type, and route time of administration of the chelator, and whether the calcium or zinc salt of DTPA is used. Current and future research efforts should be directed at overcoming limitations associated with the use of these complex drugs by using innovative methods that can enhance their structural and therapeutic properties.

  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. 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

  5. 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...

  6. 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)

  7. 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.

  8. 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)

  9. 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.

  10. 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)

  11. 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

  12. 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.)

  13. 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

  14. 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)

  15. 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.)

  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. 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

  18. EFFECT OF COMPOSITION OF SELECTED GROUNDWATERS FROM THE BASIN AND RANGE PROVINCE ON PLUTONIUM, NEPTUNIUM, AND AMERICIUM SPECIATION.

    Science.gov (United States)

    Rees, Terry F.; Cleveland, Jess M.; Nash, Kenneth L.

    1984-01-01

    The speciation of plutonium, neptunium, and americium was determined in groundwaters from four sources in the Basin and Range Province: the lower carbonate aquifer, Nevada Test Site (NTS) (Crystal Pool); alluvial fill, Frenchman Flat, NTS (well 5C); Hualapai Valley, Arizona (Red Lake south well); and Tularosa Basin, New Mexico (Rentfrow well). The results were interpreted to indicate that plutonium and, to a lesser extent, neptunium are least soluble in reducing groundwaters containing a large concentration of sulfate ion and a small concentration of strongly complexing anions. The results further emphasize the desirability of including studies such as this among the other site-selection criteria for nuclear waste repositories.

  19. 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

  20. 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).

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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.)

  9. 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.

  10. 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

  11. 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

  12. 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)

  13. 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...

  14. 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)

  15. 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.

  16. 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%

  17. 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)

  18. 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)

  19. 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

  20. 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)

  1. A Graphical Examination of Uranium and Plutonium Fissility

    Science.gov (United States)

    Reed, B. Cameron

    2008-01-01

    The issue of why only particular isotopes of uranium and plutonium are suitable for use in nuclear weapons is analyzed with the aid of graphs and semiquantitative discussions of parameters such as excitation energies, fission barriers, reaction cross-sections, and the role of processes such as [alpha]-decay and spontaneous fission. The goal is to…

  2. 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)

  3. 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)

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

    Science.gov (United States)

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

    2015-09-01

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

  5. 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.)

  6. 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)

  7. 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

  8. 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

  9. 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)

  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. 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

  12. 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

  13. Long-term logistic analysis of FBR introduction strategy: avoiding both uranium and plutonium shortage

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, T. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Center for International Studies

    1995-12-31

    Despite comfortable predictions on short to mid-term uranium resources, there is still a concern about long-term availability of competitive uranium resources. In order to achieve substantial uranium saving, early introduction of Fast Breeder Reactor (FBR) is desirable. But it is also known that rapid introduction of FBR could result in plutonium storage. Will there be enough plutonium on a global scale to sustain fast FBR growth? is there any other way to save uranium resource? This paper concludes that multi-option strategies to achieve flexible long-term strategy to avoid both uranium and plutonium storage are desirable. (authors) 3 refs.

  14. 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,

  15. 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)

  16. 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)

  17. 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.)

  18. 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

  19. 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

  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. 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)

  2. Americium-241 and plutonium-237 turnover in mussels ( Mytilus galloprovincialis) living in field enclosures

    Science.gov (United States)

    Guary, J. C.; Fowler, S. W.

    1981-02-01

    Loss of 241Am and 237Pu from contaminated mussels ( Mytilus galloprovincialis) living in situ in the Mediterranean Sea is described as the sum of three exponential functions. In the case of 241Am, two short-lived compartments representing a total of 80% of the incorporated radionuclide turned over rapidly with biological half-lives of 2 and 3 weeks. The remaining fraction of 241Am, associated with a long-lived compartment, was lost at an extremely slow rate ( Tb1/2=1·3 years). Plutonium-237 turnover in the two short-lived compartments (containing 70% of the Pu) was more rapid ( Tb1/2=1-2 days and 2 weeks) than that of 241Am; however, there was some indication that subsequent loss rates of the two radionuclides in long-lived compartments may be similar if determined over comparable periods of time. Loss rates of 241Am differed for the various tissues, with the most rapid rates occurring in gill, viscera and shell. Abrupt changes in loss observed in muscle and mantle suggested a translocation of 241Am to muscle and mantle during depuration. Whole shell contained by far the largest fraction (˜90%) of both 241Am and 237Pu taken up; in addition, these radionuclides are not irreversibly bound to mussel shell but readily leach into the water. These observations suggest that mollusc shell may influence the biogeochemistry of transuranic elements in littoral zones.

  3. 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)

  4. 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)

  5. Transfer across the human gut of environmental plutonium, americium, cobalt, caesium and technetium: studies with cockles (Cerastoderma edule) from the Irish Sea.

    Science.gov (United States)

    Hunt, G J

    1998-06-01

    Our previous studies have indicated lower values of the gut transfer factor ('f1 values') for plutonium and americium in winkles (Littorina littorea) than adopted by ICRP. The present study was undertaken primarily to investigate whether this observation extends to other species. Samples of cockles (Carastoderma edule) from Ravenglass, Cumbria were eaten by volunteers who provided 24 h samples of urine and faeces. Urine samples indicated f1 values for cockles which were higher than for winkles; for plutonium these ranged overall up to 7 x 10(-4) with an arithmetic mean in the range (2-3) x 10(-4), and for americium up to 2.6 x 10(-4) with an arithmetic mean of 1.2 x 10(-4). Limited data based on volunteers eating cockles from the Solway suggest that f1 values for americium may be greater at distance from Sellafield. The measured values compare with 5 x 10(-4) used by the ICRP for environmental forms of both elements, which would appear to provide adequate protection when calculating doses from Cumbrian cockles. Data for other nuclides were obtained by analysing faecal samples from the volunteers who ate the Ravenglass cockles. Cobalt-60 showed an f1 value in the region of 0.2, twice the value currently used by ICRP. For 137Cs, variabilities were indicated in the range 0.08 to 0.43, within the ICRP value of f1 = 1.0. Technetium-99 gave f1 values up to about 0.6, in reasonable conformity with the ICRP value of 0.5.

  6. Transfer across the human gut of environmental plutonium, americium, cobalt, caesium and technetium: studies with cockles (Cerastoderma edule) from the Irish Sea

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, G.J. [CEFAS Laboratory, Lowestoft, Suffolk NR33 0HT (United Kingdom)

    1998-06-01

    Our previous studies have indicated lower values of the gut transfer factor ('f{sub L} values') for plutonium and americium in winkles (Littorina littorea) than adopted by ICRP. The present study was undertaken primarily to investigate whether this observation extends to other species. Samples of cockles (Cerastoderma edule) from Ravenglass, Cumbria were eaten by volunteers who provided 24 samples of urine and faeces. Urine samples indicated f{sub L} values for cockles which were higher than for winkles; for plutonium these ranged overall up to 7x10{sup -4} with an arithmetic mean in the range (2-3)x10{sup -4}, and for americium up to 2.6x10{sup -4} with an arithmetic mean of 1.2x10{sup -4}. Limited data based on volunteers eating cockles from the Solway suggest that f{sub L} values for americium may be greater at distance from Sellafield. The measured values compare with 5x10{sup -4} used by the ICRP for environmental forms of both elements, which would appear to provide adequate protection when calculating doses from Cumbrian cockles. Data for other nuclides were obtained by analysing faecal samples from the volunteers who ate the Ravenglass cockles. Cobalt-60 showed an f{sub L} value in the region of 0.2, twice the value currently used by ICRP. For {sup 137}Cs, variabilities were indicated in the range 0.08 to 0.43, within the ICRP value of f{sub L}=1.0. Technetium-99 gave f{sub L} values up to about 0.6, in reasonable conformity with the ICRP value of 0.5. (author)

  7. 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.

  8. 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

  9. 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

  10. 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

  11. Reactions of plutonium dioxide with water and oxygen-hydrogen mixtures: Mechanisms for corrosion of uranium and plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Haschke, John M.; Allen, Thomas H.; Morales, Luis A.

    1999-06-18

    Investigation of the interactions of plutonium dioxide with water vapor and with an oxygen-hydrogen mixture show that the oxide is both chemically reactive and catalytically active. Correspondence of the chemical behavior with that for oxidation of uranium in moist air suggests that similar catalytic processes participate in the mechanism of moisture-enhanced corrosion of uranium and plutonium. Evaluation of chemical and kinetic data for corrosion of the metals leads to a comprehensive mechanism for corrosion in dry air, water vapor, and moist air. Results are applied in confirming that the corrosion rate of Pu in water vapor decreases sharply between 100 and 200 degrees C.

  12. Electrolytic partitioning of uranium and plutonium based on a new type of electrolytic mixer-settler

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Z.; Yan, T.; Zheng, W. [China Institute of Atomic Energy, Beijing (China)] [and others

    2013-11-01

    The design of 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 described and the operational results achieved using this equipment are presented. The results show that this new type of mixer-settler has excellent separation performances. The flow rate of organic feed solution is 3 mL/min and the flow ratio of feed solution (1BF) to aqueous back extraction stream (1BX) and to organic wash stream (1BS) is 4/1/1. For an organic feed of 84 g/L uranium and 1.40 {proportional_to} 2.64 g/L plutonium, both the separation factor of plutonium from uranium and that of uranium from plutonium are apparently higher than 10{sup 4}. (orig.)

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. Plutonium

    Science.gov (United States)

    Clark, David L.; Hecker, Siegfried S.; Jarvinen, Gordon D.; Neu, Mary P.

    The element plutonium occupies a unique place in the history of chemistry, physics, technology, and international relations. After the initial discovery based on submicrogram amounts, it is now generated by transmutation of uranium in nuclear reactors on a large scale, and has been separated in ton quantities in large industrial facilities. The intense interest in plutonium resulted fromthe dual-use scenario of domestic power production and nuclear weapons - drawing energy from an atomic nucleus that can produce a factor of millions in energy output relative to chemical energy sources. Indeed, within 5 years of its original synthesis, the primary use of plutonium was for the release of nuclear energy in weapons of unprecedented power, and it seemed that the new element might lead the human race to the brink of self-annihilation. Instead, it has forced the human race to govern itself without resorting to nuclear war over the past 60 years. Plutonium evokes the entire gamut of human emotions, from good to evil, from hope to despair, from the salvation of humanity to its utter destruction. There is no other element in the periodic table that has had such a profound impact on the consciousness of mankind.

  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. Nano-cerium vanadate: a novel inorganic ion exchanger for removal of americium and uranium from simulated aqueous nuclear waste.

    Science.gov (United States)

    Banerjee, Chayan; Dudwadkar, Nilesh; Tripathi, Subhash Chandra; Gandhi, Pritam Maniklal; Grover, Vinita; Kaushik, Chetan Prakash; Tyagi, Avesh Kumar

    2014-09-15

    Cerium vanadate nanopowders were synthesized by a facile low temperature co-precipitation method. The product was characterized by X-ray diffraction and transmission electron microscopy and found to consist of ∼25 nm spherical nanoparticles. The efficiency of these nanopowders for uptake of alpha-emitting radionuclides (233)U (4.82 MeV α) and (241)Am (5.49 MeV α, 60 keV γ) has been investigated. Thermodynamically and kinetically favorable uptake of these radionuclides resulted in their complete removal within 3h from aqueous acidic feed solutions. The uptake capacity was observed to increase with increase in pH as the zeta potential value decreased with the increase in pH but effect of ionic strength was insignificant. Little influence of the ions like Sr(2+), Ru(3+), Fe(3+), etc., in the uptake process indicated CeVO4 nanopowders to be amenable for practical applications. The isotherms indicated predominant uptake of the radioactive metal ions in the solid phase of the exchanger at lower feed concentrations and linear Kielland plots with positive slopes indicated favorable exchange of the metal ions with the nanopowder. Performance comparison with the other sorbents reported indicated excellent potential of nano-cerium vanadate for removing americium and uranium from large volumes of aqueous acidic solutions.

  1. 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)

  2. 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.)

  3. 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

  4. 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

  5. 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.

  6. 10 CFR Appendix J to Part 110 - Illustrative List of Uranium Conversion Plant Equipment and Plutonium Conversion Plant Equipment...

    Science.gov (United States)

    2010-01-01

    ... and Plutonium Conversion Plant Equipment Under NRC Export Licensing Authority J Appendix J to Part 110... Plutonium Conversion Plant Equipment Under NRC Export Licensing Authority Note—Uranium conversion plants and... feed for electromagnetic enrichment. Note: Plutonium conversion plants and systems may perform one...

  7. 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

  8. METHOD OF SEPARATING URANIUM, PLUTONIUM AND FISSION PRODUCTS BY BROMINATION AND DISTILLATION

    Science.gov (United States)

    Jaffey, A.H.; Seaborg, G.T.

    1958-12-23

    The method for separation of plutonium from uranium and radioactive fission products obtained by neutron irradiation of uranlum consists of reacting the lrradiated material with either bromine, hydrogen bromide, alumlnum bromide, or sulfur and bromine at an elevated temperature to form the bromides of all the elements, then recovering substantlally pure plutonium bromide by dlstillatlon in combinatlon with selective condensatlon at prescribed temperature and pressure.

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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...

  14. 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.)

  15. Standard test method for uranium and plutonium concentrations and isotopic abundances by thermal ionization mass spectrometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This test method covers the determination of the concentration and isotopic composition of uranium and plutonium in solutions. The purified uranium or plutonium from samples ranging from nuclear materials to environmental or bioassay matrices is loaded onto a mass spectrometric filament. The isotopic ratio is determined by thermal ionization mass spectrometry, the concentration is determined by isotope dilution. 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 safety and health practices and determine the applicability of regulatory limitations prior to use.

  16. 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)

  17. 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

  18. 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.

  19. Assessing alternative strategies for the disposition of weapons-grade uranium and plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Chow, B.G.

    1995-12-31

    Highly-enriched uranium (HEU) from dismantled nuclear weapons and military inventory can be blended down into proliferation-resistant low-enriched uranium and used economically as fuel in current nuclear reactors. However, the US can no longer expect the agreement to purchase and resell the uranium blended down from 500 metric tons of Russia`s HEU to be budget neutral. The authors recommend that other countries participate in the repurchase of blended-down uranium from the US and that a multilateral offer to Russia, which acts on behalf of all four former Soviet nuclear republics, be made for the purchase of the blended-down uranium from Russia`s remaining HEU. Since spent fuel in temporary storage worldwide contains enough plutonium to fuel breeders on any realistic buildup schedule in the event that breeders are needed, there is no need to save the weapons-grade plutonium for the future. This paper compares the costs of burning it in existing light water reactors, storing it indefinitely, and burying it after 20 years of storage. They found that the present-valued cost is about $1 to 2 billion in US dollars for all three alternatives. The deciding factor for selection should be an alternative`s proliferation resistance. Prolonged plutonium storage in Russia runs the risk of theft and, if the Russian political scene turns for the worse, the risk of re-use in its nuclear arsenal. The most urgent issue, however, is to determine not the disposition alternative but whether Russia will let its weapons-grade plutonium leave the former Soviet Republics (FSRs). The US should offer to buy and remove such plutonium from the FSRs. If Russia refuses even after the best US efforts, the US should then persuade Russia to burn or bury the plutonium, but not store it indefinitely for future breeder use.

  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. Nano-cerium vanadate: A novel inorganic ion exchanger for removal of americium and uranium from simulated aqueous nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, Chayan; Dudwadkar, Nilesh [Fuel Reprocessing Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Tripathi, Subhash Chandra, E-mail: sctri001@gmail.com [Fuel Reprocessing Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gandhi, Pritam Maniklal [Fuel Reprocessing Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Grover, Vinita [Waste Management Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kaushik, Chetan Prakash [Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Tyagi, Avesh Kumar, E-mail: aktyagi@barc.gov.in [Waste Management Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-09-15

    Highlights: • Template free, low temperature synthesis of CeVO{sub 4} nanopowders. • Thermodynamically and kinetically favourable uptake of Am(III) and U(VI) exhibited. • K{sub d} and ΔG° values for Am(III) and U(VI) uptake in pH 1–6 are reported. • Interdiffusion coefficients and zeta potential values in pH 1–6 are reported. • Possible application in low level aqueous nuclear waste remediation. - Abstract: Cerium vanadate nanopowders were synthesized by a facile low temperature co-precipitation method. The product was characterized by X-ray diffraction and transmission electron microscopy and found to consist of ∼25 nm spherical nanoparticles. The efficiency of these nanopowders for uptake of alpha-emitting radionuclides {sup 233}U (4.82 MeV α) and {sup 241}Am (5.49 MeV α, 60 keV γ) has been investigated. Thermodynamically and kinetically favorable uptake of these radionuclides resulted in their complete removal within 3 h from aqueous acidic feed solutions. The uptake capacity was observed to increase with increase in pH as the zeta potential value decreased with the increase in pH but effect of ionic strength was insignificant. Little influence of the ions like Sr{sup 2+}, Ru{sup 3+}, Fe{sup 3+}, etc., in the uptake process indicated CeVO{sub 4} nanopowders to be amenable for practical applications. The isotherms indicated predominant uptake of the radioactive metal ions in the solid phase of the exchanger at lower feed concentrations and linear Kielland plots with positive slopes indicated favorable exchange of the metal ions with the nanopowder. Performance comparison with the other sorbents reported indicated excellent potential of nano-cerium vanadate for removing americium and uranium from large volumes of aqueous acidic solutions.

  2. 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)

  3. 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)

  4. 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).

  5. 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).

  6. 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.

  7. Production and characterization of monodisperse plutonium, uranium, and mixed uranium-plutonium particles for nuclear safeguard applications.

    Science.gov (United States)

    Ranebo, Y; Niagolova, N; Erdmann, N; Eriksson, M; Tamborini, G; Betti, M

    2010-05-15

    In order to prevent nuclear proliferation, the isotopic analysis of uranium and plutonium microparticles has strengthened the means in international safeguards for detecting undeclared nuclear activities. In order to ensure accuracy and precision in the analytical methodologies used, the instrumental techniques need to be calibrated. The objective of this study was to produce and characterize particles consisting of U, Pu, and mixed U-Pu, suitable for such reliability verifications. A TSI vibrating orifice aerosol generator in connection with a furnace system was used to produce micrometer sized, monodispersed particles from reference U and Pu materials in solution. The particle masses (in the range of 3-6 pg) and sizes (approximately 1.5 microm) were controlled by the experimental conditions and the parameters for the aerosol generator. Size distributions were obtained from scanning electron microscopy, and energy-dispersive X-ray analysis confirmed that the particle composition agreed with the starting material used. A secondary ion mass spectrometer (SIMS) was used to characterize the isotopic composition of the particles. Isobaric and polyatomic interference in the SIMS spectra was identified. In order to obtain accurate estimates of the interference, a batch of Pu particles were produced of mainly (242)Pu. These were used for SIMS analysis to characterize the relative ionization of Pu and U hydride ions and to determine the SIMS useful yields of U and Pu. It was found that U had a higher propensity to form the hydride than Pu. Useful yields were determined at a mass resolution of 450 for U-Pu particles: (1.71 +/- 0.15) % for Pu and (0.72 +/- 0.06) % for U. For Pu particles: (1.65 +/- 0.14) % for Pu. This gave a relative sensitivity factor between U and Pu (RSF(U:Pu)) of 2.4 +/- 0.2. However, the RSF(U:Pu) showed large fluctuations during the sputtering time for each analyses of the mixed U-Pu particles, in the range of 1.9-3.4. PMID:20415437

  8. 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)

  9. 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.

  10. 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

  11. 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.

  12. High temperature radiance spectroscopy measurements of solid and liquid uranium and plutonium carbides

    Science.gov (United States)

    Manara, D.; De Bruycker, F.; Boboridis, K.; Tougait, O.; Eloirdi, R.; Malki, M.

    2012-07-01

    In this work, an experimental study of the radiance of liquid and solid uranium and plutonium carbides at wavelengths 550 nm ⩽ λ ⩽ 920 nm is reported. A fast multi-channel spectro-pyrometer has been employed for the radiance measurements of samples heated up to and beyond their melting point by laser irradiation. The melting temperature of uranium monocarbide, soundly established at 2780 K, has been taken as a radiance reference. Based on it, a wavelength-dependence has been obtained for the high-temperature spectral emissivity of some uranium carbides (1 ⩽ C/U ⩽ 2). Similarly, the peritectic temperature of plutonium monocarbide (1900 K) has been used as a reference for plutonium monocarbide and sesquicarbide. The present spectral emissivities of solid uranium and plutonium carbides are close to 0.5 at 650 nm, in agreement with previous literature values. However, their high temperature behaviour, values in the liquid, and carbon-content and wavelength dependencies in the visible-near infrared range have been determined here for the first time. Liquid uranium carbide seems to interact with electromagnetic radiation in a more metallic way than does the solid, whereas a similar effect has not been observed for plutonium carbides. The current emissivity values have also been used to convert the measured radiance spectra into real temperature, and thus perform a thermal analysis of the laser heated samples. Some high-temperature phase boundaries in the systems U-C and Pu-C are shortly discussed on the basis of the current results.

  13. 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

  14. 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

  15. 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)

  16. 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

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

    Science.gov (United States)

    Fisenko, Anatoliy I.; Lemberg, Vladimir F.

    2016-09-01

    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 functions have not been established for plutonium carbide. The calculated values for the normal total emissivity of uranium and plutonium carbides at their melting temperatures is in good agreement with experimental data. The obtained results allow to calculate the thermal radiative and thermodynamic properties of liquid and solid uranium and plutonium carbides for any size of samples. Based on the model of Hagen-Rubens and the Wiedemann-Franz law, a new method to determine the thermal conductivity of metals and carbides at the melting points is proposed.

  18. 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)

  19. 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)

  20. Standard test method for atom percent fission in uranium and plutonium fuel (Neodymium-148 Method)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1996-01-01

    1.1 This test method covers the determination of stable fission product 148Nd in irradiated uranium (U) fuel (with initial plutonium (Pu) content from 0 to 50 %) as a measure of fuel burnup (1-3). 1.2 It is possible to obtain additional information about the uranium and plutonium concentrations and isotopic abundances on the same sample taken for burnup analysis. If this additional information is desired, it can be obtained by precisely measuring the spike and sample volumes and following the instructions in Test Method E267. 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.

  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. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. Polonium, uranium and plutonium radionuclides in aquatic and land ecosystem of Poland.

    Science.gov (United States)

    Skwarzec, Bogdan; Strumińska-Parulska, Dagmara I; Boryło, Alicja; Kabat, Krzysztof

    2012-01-01

    This article presents the results of study about distribution, inflow and accumulation of polonium, uranium and plutonium in aquatic and land environment of Poland and the southern Baltic Sea. Radionuclides of (210)Po, (234)U and (238)U as well as (239+240)Pu and (241)Pu are strongly accumulated in Baltic organisms and plants and transferred through the trophic chain. The values of bioconcentration factor (BCF) in Baltic plants and animals are higher for polonium and plutonium in comparison with uranium. The principal source of radionuclides in the southern Baltic Sea is their inflow with rivers. Total annual runoff of polonium, uranium and plutonium from the Vistula and the Odra as well as the Pomeranian rivers were calculated at 95 GBq of (210)Po, 750 GBq of (234+238)U and 160 MBq of (238+239+240)Pu. Seasonal and spatial variability of (210)Po, (238)U and (239+240)Pu levels in the Vistula and the Odra drainage basins were assessed by application of neural-network based classification, especially cluster analysis (CA), principal component analysis (PCA) and self-organizing maps (SOM). The result for the Vistula river indicated correlation between polonium and plutonium as well as polonium and uranium. In the Odra drainage basin, the biggest differences were observed in the case of (238)U. To assess if there are statistically significant differences in mean concentration values of (210)Po, (238)U and (239+240)Pu for the Vistula and the Odra rivers drainage basins were obtained by used of the non-parametric tests. Comparing to the Vistula catchment area, statistically differences concentration of (210)Po and (239+240)Pu in all year was observed for river samples collected on the Odra drainage basin.

  8. 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)

  9. 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.

  10. 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.

  11. 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

  12. 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 ...

  13. 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.)

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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.

  19. 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

  20. 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.

  1. 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)

  2. 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...

  3. 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.

  4. 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)

  5. 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)

  6. 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.

  7. Simple calculation of the critical mass for highly enriched uranium and plutonium-239

    Science.gov (United States)

    Chyba, Christopher F.; Milne, Caroline R.

    2014-10-01

    The correct calculation of and values for the critical mass of uranium or plutonium necessary for a nuclear fission weapon have long been understood and publicly available. The calculation requires solving the radial component in spherical coordinates of a diffusion equation with a source term, so is beyond the reach of most public policy and many first-year college physics students. Yet it is important for the basic physical ideas behind the calculation to be understood by those without calculus who are nonetheless interested in international security, arms control, or nuclear non-proliferation. This article estimates the critical mass in an intuitive way that requires only algebra.

  8. 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

  9. 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

  10. 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)

  11. 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.)

  12. 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)

  13. 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.)

  14. 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.

  15. 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)

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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

  8. Plutonium and Uranium Atom Ratios and Activity Levels in Cochiti Lake Bottom Sediments Provided by Pueblo de Cochiti

    Energy Technology Data Exchange (ETDEWEB)

    Gallaher, B.M.; Efurd, D.W.; Rokop, D.J.; Benjamin, T.M.

    1999-05-01

    Historical operations at the Los Alamos National Laboratory have contaminated stream sediments with plutonium and other radionuclides. A small portion of these contaminated sediments has been carried by floods into the Rio Grande drainage system, eventually to be trapped by Cochiti Lake located on Pueblo de Cochiti lands approximately 8 km downstream of the Laboratory. In this study, lake bottom sediment samples provided by the Pueblo de Cochiti were analyzed by thermal ionization mass spectrometry to determine plutonium and uranium activity levels and isotopic atom ratios. This specialized analytical method allows us to take isotopic fingerprints of radionuclides found in the sediment and to determine how much plutonium and uranium came from the Laboratory and how much was deposited by worldwide fallout or is natural. Two distinct types of samples were processed: segments of a continuous vertical core of the entire accumulated sediment sequence and other samples from across the lake bottom at the water/sediment interface. Based on measurement of the {sup 240}Pu/{sup 239}Pu atom ratio, Laboratory-derived plutonium is present in eight of nine samples at the core site. On a depth-weighted basis, approximately one-half of the {sup 239}Pu and {sup 240}Pu came from early operations at the Laboratory; the remaining plutonium came from fallout dispersed by above-ground nuclear tests. In contrast to the core site, the samples from the other locations showed little or no evidence of Laboratory-derived plutonium, with more than 90 percent of the plutonium attributable to fallout. The overall amount of plutonium in all the samples is of the same magnitude as other reservoirs in the region. The net increase in plutonium over upstream reservoirs unaffected by Laboratory activities is a maximum of 0.014 pCi/g or 3.5 times. All of the samples reflect natural uranium compositions. Laboratory-derived uranium is not identifiable, presumably because the sediment contains abundant

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Scherer, Carolynn P [Los Alamos National Laboratory; Long, Jon D [Los Alamos National Laboratory

    2010-09-28

    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

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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.

  16. 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

  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. 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.).

  19. 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.

  20. 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

  1. 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)

  2. 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

  3. 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.

  4. Late-occurring pulmonary pathologies following inhalation of mixed oxide (uranium + plutonium oxide) aerosol in the rat.

    Science.gov (United States)

    Griffiths, N M; Van der Meeren, A; Fritsch, P; Abram, M-C; Bernaudin, J-F; Poncy, J L

    2010-09-01

    Accidental exposure by inhalation to alpha-emitting particles from mixed oxide (MOX: uranium and plutonium oxide) fuels is a potential long-term health risk to workers in nuclear fuel fabrication plants. For MOX fuels, the risk of lung cancer development may be different from that assigned to individual components (plutonium, uranium) given different physico-chemical characteristics. The objective of this study was to investigate late effects in rat lungs following inhalation of MOX aerosols of similar particle size containing 2.5 or 7.1% plutonium. Conscious rats were exposed to MOX aerosols and kept for their entire lifespan. Different initial lung burdens (ILBs) were obtained using different amounts of MOX. Lung total alpha activity was determined by external counting and at autopsy for total lung dose calculation. Fixed lung tissue was used for anatomopathological, autoradiographical, and immunohistochemical analyses. Inhalation of MOX at ILBs ranging from 1-20 kBq resulted in lung pathologies (90% of rats) including fibrosis (70%) and malignant lung tumors (45%). High ILBs (4-20 kBq) resulted in reduced survival time (N = 102; p plutonium oxide alone (1.9% Gy). Staining with antibodies against Surfactant Protein-C, Thyroid Transcription Factor-1, or Oct-4 showed differential labeling of tumor types. In conclusion, late effects following MOX inhalation result in similar risk for development of lung tumors as compared with industrial plutonium oxide.

  5. 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

  6. 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.

  7. 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.

  8. 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

  9. 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)

  10. 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)

  11. 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)

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. Isotope ratio analysis of individual sub-micrometer plutonium particles with inductively coupled plasma mass spectrometry.

    Science.gov (United States)

    Esaka, Fumitaka; Magara, Masaaki; Suzuki, Daisuke; Miyamoto, Yutaka; Lee, Chi-Gyu; Kimura, Takaumi

    2010-12-15

    Information on plutonium isotope ratios in individual particles is of great importance for nuclear safeguards, nuclear forensics and so on. Although secondary ion mass spectrometry (SIMS) is successfully utilized for the analysis of individual uranium particles, the isobaric interference of americium-241 to plutonium-241 makes difficult to obtain accurate isotope ratios in individual plutonium particles. In the present work, an analytical technique by a combination of chemical separation and inductively coupled plasma mass spectrometry (ICP-MS) is developed and applied to isotope ratio analysis of individual sub-micrometer plutonium particles. The ICP-MS results for individual plutonium particles prepared from a standard reference material (NBL SRM-947) indicate that the use of a desolvation system for sample introduction improves the precision of isotope ratios. In addition, the accuracy of the (241)Pu/(239)Pu isotope ratio is much improved, owing to the chemical separation of plutonium and americium. In conclusion, the performance of the proposed ICP-MS technique is sufficient for the analysis of individual plutonium particles. PMID:21111176

  18. Isotope ratio analysis of individual sub-micrometer plutonium particles with inductively coupled plasma mass spectrometry.

    Science.gov (United States)

    Esaka, Fumitaka; Magara, Masaaki; Suzuki, Daisuke; Miyamoto, Yutaka; Lee, Chi-Gyu; Kimura, Takaumi

    2010-12-15

    Information on plutonium isotope ratios in individual particles is of great importance for nuclear safeguards, nuclear forensics and so on. Although secondary ion mass spectrometry (SIMS) is successfully utilized for the analysis of individual uranium particles, the isobaric interference of americium-241 to plutonium-241 makes difficult to obtain accurate isotope ratios in individual plutonium particles. In the present work, an analytical technique by a combination of chemical separation and inductively coupled plasma mass spectrometry (ICP-MS) is developed and applied to isotope ratio analysis of individual sub-micrometer plutonium particles. The ICP-MS results for individual plutonium particles prepared from a standard reference material (NBL SRM-947) indicate that the use of a desolvation system for sample introduction improves the precision of isotope ratios. In addition, the accuracy of the (241)Pu/(239)Pu isotope ratio is much improved, owing to the chemical separation of plutonium and americium. In conclusion, the performance of the proposed ICP-MS technique is sufficient for the analysis of individual plutonium particles.

  19. 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.

  20. 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)

  1. 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)

  2. 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)

  3. 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.)

  4. 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...

  5. 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)

  6. 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)

  7. 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.

  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. 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.

  10. 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

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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)

  17. 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

  18. 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.)

  19. 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

  20. 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)

  1. 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

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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.)

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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)

  14. 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)

  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. 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)

  17. 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)

  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. 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.

  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

    1992-08-01

    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.

  2. 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

  3. 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

  4. 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

  5. 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...

  6. 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)

  7. 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)

  8. 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)

  9. Optimization of enrichment distributions in nuclear fuel assemblies loaded with uranium and plutonium via a modified linear programming technique

    Science.gov (United States)

    Cuevas Vivas, Gabriel Francisco

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed OXide (MOX) fuels. MOX isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide-range of applicability of the optimization technique. The features of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or

  10. 1976 Hanford americium accident

    Energy Technology Data Exchange (ETDEWEB)

    Heid, K R; Breitenstein, B D; Palmer, H E; McMurray, B J; Wald, N

    1979-01-01

    This report presents the 2.5-year medical course of a 64-year-old Hanford nuclear chemical operator who was involved in an accident in an americium recovery facility in August 1976. He was heavily externally contaminated with americium, sustained a substantial internal deposition of this isotope, and was burned with concentrated nitric acid and injured by flying debris about the face and neck. The medical care given the patient, including the decontamination efforts and clinical laboratory studies, are discussed. In-vivo measurements were used to estimate the dose rates and the accumulated doses to body organs. Urinary and fecal excreta were collected and analyzed for americium content. Interpretation of these data was complicated by the fact that the intake resulted both from inhalation and from solubilization of the americium embedded in facial tissues. A total of 1100 ..mu..Ci was excreted in urine and feces during the first 2 years following the accident. The long-term use of diethylenetriaminepentate (DTPA), used principally as the zinc salt, is discussed including the method, route of administration, and effectiveness. To date, the patient has apparently experienced no complications attributable to this extensive course of therapy, even though he has been given approximately 560 grams of DTPA. 4 figures, 1 table.

  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. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

  13. 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

  14. 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

  15. 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)

  16. The IDA-80 measurement evaluation programme on mass spectrometric isotope dilution analysis of uranium and plutonium. Vol. 1

    International Nuclear Information System (INIS)

    The main objective was the acquisition of basic data on the uncertainties involved in the mass spectrometric isotope dilution analysis as applied to the determination of uranium and plutonium in active feed solutions of reprocessing plants. The element concentrations and isotopic compositions of all test materials used were determined by CBNM and NBS with high accuracy. The more than 60000 analytical data reported by the participating laboratories were evaluated by statistical methods applied mainly to the calculation of estimates of the variances for the different uncertainty components contributing to the total uncertainty of this analytical technique. Attention was given to such topics as sample ageing, influence of fission products, spike calibration, ion fractionation, Pu-241 decay correction, minor isotope measurement and errors in data transfer. Furthermore, the performance of the 'dried sample' technique and the 'in-situ' spiking method of undiluted samples of reprocessing fuel solution with U-235/Pu-242 metal alloy spikes, were tested successfully. Considerable improvement of isotope dilution analysis in this safeguards relevant application during the last decade is shown as compared to the results obtained in the IDA-72 interlaboratory experiment, organized by KfK in 1972 on the same subject. (orig./HP)

  17. 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.

  18. 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.

  19. 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.

  20. Uranium

    International Nuclear Information System (INIS)

    The article includes a historical preface about uranium, discovery of portability of sequential fission of uranium, uranium existence, basic raw materials, secondary raw materials, uranium's physical and chemical properties, uranium extraction, nuclear fuel cycle, logistics and estimation of the amount of uranium reserves, producing countries of concentrated uranium oxides and percentage of the world's total production, civilian and military uses of uranium. The use of depleted uranium in the Gulf War, the Balkans and Iraq has caused political and environmental effects which are complex, raising problems and questions about the effects that nuclear compounds left on human health and environment.

  1. Seismic tests of gloveboxes and equipment for plutonium-uranium mixed oxide fuel fabrication

    International Nuclear Information System (INIS)

    Although the seismic calculation for the gloveboxes in the plutonium fuel fabrication facilities has been carried out, the seismic test has not yet been performed. Therefore, a series of the seismic test of gloveboxes and the equipments in them was carried out. The purpose of the test was as follows. Investigation of the dynamic motion of gloveboxes and the equipments in them when an earthquake occurs, investigation of the confining ability of gloveboxes after an earthquake occurs, and investigation of the reliability of the analytical procedure by comparing the results of seismic test with those of the dynamic analysis. For the test, the upper glovebox, the lower glovebox and the model frame supporting the upper glovebox were used. The material transporting equipment was in the upper glovebox, and the pressing equipment was in the lower glovebox. Impulse test, free vibration test, sweep test, step by step test, earthquake test and leak test were carried out. The natural frequency and damping coefficient were determined, and the maximum mode of response acceleration to earthquake waves was obtained. In the leak test, the gloveboxes were normal. Generally the calculated values agreed with the experimental values. (Kako, I.)

  2. Glass-Ceramic Waste Forms for Uranium and Plutonium Residues Wastes - 13164

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, Martin W.A.; Moricca, Sam A.; Zhang, Yingjie; Day, R. Arthur; Begg, Bruce D. [Australian Nuclear Science and Technology Organisation (ANSTO), New Illawarra Road, Lucas Heights, NSW 2234 (Australia); Scales, Charlie R.; Maddrell, Ewan R. [National Nuclear Laboratory, Sellafield, Seascale, Cumbria, UK, CA20 1PG (United Kingdom); Hobbs, Jeff [Sellafield Limited, Sellafield, Seascale, Cumbria, UK, CA20 1PG (United Kingdom)

    2013-07-01

    A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibility of the glass-ceramic. (authors)

  3. The Complete Burning of Weapons Grade Plutonium and Highly Enriched Uranium with (Laser Inertial Fusion-Fission Energy) LIFE Engine

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J C; Diaz de la Rubia, T; Moses, E

    2008-12-23

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

  4. Preparation and study of the nitrides and mixed carbide-nitrides of uranium and of plutonium; Preparation et etude des nitrures et carbonitrures d'uranium et de plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Anselin, F. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1966-06-01

    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) [French] On decrit en detail une methode simple de preparation des nitrures d'uranium et de plutonium par action directe de l'azote sous pression, a temperature moyenne (vers 400 C), sur les metaux massifs partiellement hydrures. On montre que la miscibilite est complete entre les phases UN et PuN. L'evolution des parametres reticulaires des echantillons en fonction de la temperature et en presence d'oxyde a ete utilisee pour detecter et estimer la solubilite de l'oxygene dans les diverses phases. On a etudie le frittage de ces nitrures en fonction des conditions de preparation

  5. Interspecies comparison of the metabolism and dosimetry of inhaled mixed oxides of plutonium and uranium

    International Nuclear Information System (INIS)

    Three studies were conducted to provide information on the biological fate, distribution of radiation doses among tissues, and implications for potential health consequences of an inhalation exposure to mixed-oxide nuclear fuel materials. In each study, Fischer-344 rats, beagle dogs, and cynomolgus monkeys inhaled one of three aerosols: 750 degrees C calcined mixed oxides of UO2 and PuO2, 1750 degrees C sintered (U,Pu)O2, or 850 degrees C calcined open-quotes pureclose quotes PuO2. These materials were collected from glove-box enclosures immediately after industrial processing of mixed-oxide fuel materials. Lung retention, tissue distribution, and mode of excretion of 238-240Pu, 241Am, and uranium (when present) were quantified by radiochemical analysis of tissue and excreta samples from animals sacrificed at selected times to 6.5 yr after inhalation exposure

  6. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium; Contribution a l'etude du monocarbure d'uranium et de plutonium 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

    1970-03-01

    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) [French] On a etudie un combustible de type carbure (U,Pu)C pour les reacteurs a neutrons rapides. Les recherches preliminaires ont porte sur le carbure (UZr)C (rapport CEA-R-3765(1)). L'addition de faibles quantites de zirconium (3 at. pour cent) au monocarbure (U,Pu)C, ameliore certaines proprietes, commee la tenue a la corrosion atmospherique, la durete et surtout la compatibilite avec l'acier inoxydable X-18 M, Par contre le coefficient de dilatation et la densite sont peu changes. Le rapport Pu/Pu+U etait fixe a 20 pour cent. Deux procedes de fabrication ont ete etudies: l'un par fusion a l'arc, l'autre par frittage a partir de metaux hydrures. Au vu des resultats metallurgiques obtenus le carbure (U,Pu,Zr)C semble presenter un interet certain. (auteur)

  7. Getting the plutonium disposition job done: the concept of a joint-venture disposition enterprise financed by additional sales of highly enriched uranium

    International Nuclear Information System (INIS)

    The paper gives an outline of a concept which has the potential to provide both substantial financing needed for disposition of plutonium from excess nuclear weapons and the long-term management structure required to implement this effort. The three most important issues were underlined. First, it is urgent to modernize security and accounting systems for all weapon-usable nuclear materials, particularly from former Soviet Union. Second, excess plutonium and Highly Enriched Uranium (HEU) must be brought under international monitoring to ensure irreversibility of nuclear arms reduction. Third, quick move should be done towards actual disposition of excess plutonium and HEU. Technology already exists, but the key issues are how to get finance and manage this operation, particularly given its immense scope and controversial nature. An international joint venture 'Enterprise for nuclear Security' that would build and operate plutonium disposition facilities under stringent non-proliferation controls, financed through additional sales of HEU is a potentially promising approach to addressing the most difficult issues facing the disposition problem

  8. Characterization of the deviation of the ideality of concentrated electrolytic solutions: plutonium 4 and uranium 4 nitrate salts study

    International Nuclear Information System (INIS)

    The purpose of this work was to establish a new binary data base by compiling the activity coefficients of plutonium and uranium at oxidation state +IV to better account for media effects in the liquid-liquid extraction operations implemented to reprocess spent nuclear fuel. Chapter 1: first reviews the basic thermodynamic concepts before describing the issues involved in acquiring binary data for the tetravalent actinides. The difficulties arise from two characteristics of this type of electrolyte: its radioactive properties (high specific activity requiring nuclearization of the experimental instrumentation) and its physicochemical properties (strong hydrolysis). After defining the notion of fictive binary data, an approach based on the thermodynamic concept of simple solutions is described in which the activity coefficient of an aqueous phase constituent is dependent on two parameters: the water activity of the system and the total concentration of dissolved constituents. The method of acquiring fictive binary electrolyte data is based on water activity measurements for ternary or quaternary actinide mixtures in nitric acid media, and binary data for nitric acid. The experimental value is then correlated with the characteristics of the fictive binary solution of the relevant electrolyte. Chapter 2: proposes more reliable binary data for nitric acid than the published equivalents, the disparities of which are discussed. The validation of the method described in Chapter 1 for acquiring fictive binary data is then addressed. The test electrolyte, for which binary data are available in the literature, is thorium(IV) nitrate. The method is validated by comparing the published binary data obtained experimentally for binary solutions with the data determined for the ternary Th(NO3)4/HNO3/H2O system investigated in this study. The very encouraging results of this comparison led us to undertake further research in this area. Chapter 3 discusses the experimental

  9. Thermal performance of fast reactor type uranium-plutonium oxide fuel pins at beginning-life conditions

    International Nuclear Information System (INIS)

    A new thermal conductivity correlation, named as ''PNC'98 equation'', of fast reactor type uranium-plutonium oxide (FR-MOX) fuel was proposed by one of the authors last year. Since maximum operating temperature of fast reactor fuel is generally much higher than that of light water reactor fuel, the predictability of the correlations for FR-MOX fuel at higher temperatures, especially, above 2300K, become more important in case of the fuel pin thermal performance analysis. Although the thermal conductivity measurements at higher temperatures are very scares, power-to-melt (PTM) experiments at beginning-of-life conditions can compensate for the lack of the measurements and be preferable to verify the correlation up to melting point. The PNC'98 equation was applied to the PTM test results irradiated in experimental fast reactor JOYO in order to examine its predictability up to melting point by integral method from a viewpoint of the heat transfer behavior across fuel-to-cladding gap through the fuel pin thermal analysis. Fuel pin irradiation behaviors, such as microstructure evolution and the porosity effect on the heat conduction in fuel pellet, were carefully modeled with some related physical properties based on the latest knowledge in order to minimize the uncertainty of the analysis. The calculated heat transfer behavior across the gap derived from temperature increments in the fuel pin is reasonably consistent to the past studies. This implies that the combinations of the PNC'98 equation with the model are highly predictable for the temperature distribution in the PTM test fuel pins and that the applicability of the PNC'98 equation is verified up to melting point. (author)

  10. Determination of uranium, thorium and plutonium isotopes by ICP-MS

    International Nuclear Information System (INIS)

    Quantitative and isotopic measurement of actinide elements is required in many circumstances in the nuclear industry. For example, determination of very low levels of these alpha emitters in human urine samples is used to assess the internal committed dose for nuclear workers. Quantifying actinide isotopes in radioactive waste from nuclear processing and nuclear facility decommissioning provides important information for waste management. Accurate determination of the uranium isotopic ratios in reactor fuels provides fuel burnup information. Inductively coupled plasma mass spectrometry (ICP-MS) has been used for the determination of Th, U, and Pu in various samples including urine, nuclear waste, and nuclear fuel in our laboratory. In order to maximize the capability of the technique and ensure quality analyses, ICP-MS was used to analyze samples directly, or after pre-treatment to separate complicated matrices or to concentrate the analyte(s). High-efficiency sample introduction techniques were investigated. Spectral interferences to minor isotopes caused by peak tails and hydride ions of major actinide isotopes were studied in detail using solutions prepared with light and heavy waters. The quality of the isotopic ratio measurement was monitored using standard reference materials. (author)

  11. Recherche de l'économie des ressources naturelles par des études de conception de coeurs de réacteurs à eau et à haut facteur de conversion à combustibles mixtes Thorium / Uranium / Plutonium

    OpenAIRE

    Vallet, Vanessa

    2012-01-01

    Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3rd 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 multirecycling strategies of both plutonium and 233U, in order to maximize natural resources economy. T...

  12. The economics of plutonium recycle

    International Nuclear Information System (INIS)

    The individual cost components and the total fuel cycle costs for natural uranium and uranium-plutonium mixed oxide fuel cycles for CANDU-PHW reactors are discussed. A calculation is performed to establish the economic conditions under which plutonium recycle would be economically attractive. (auth)

  13. Optimization of enrichment distributions in nuclear fuel assemblies loaded with Uranium and Plutonium via a modified linear programming technique

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas Vivas, Gabriel Francisco

    1999-12-01

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed Oxide (MOX) isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide range of applicability of the optimization technique. The feature of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactor (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or in the

  14. Structures of Plutonium(IV) and Uranium(VI) with N,N-Dialkyl Amides from Crystallography, X-ray Absorption Spectra, and Theoretical Calculations.

    Science.gov (United States)

    Acher, Eléonor; Hacene Cherkaski, Yanis; Dumas, Thomas; Tamain, Christelle; Guillaumont, Dominique; Boubals, Nathalie; Javierre, Guilhem; Hennig, Christoph; Solari, Pier Lorenzo; Charbonnel, Marie-Christine

    2016-06-01

    The structures of plutonium(IV) and uranium(VI) ions with a series of N,N-dialkyl amides ligands with linear and branched alkyl chains were elucidated from single-crystal X-ray diffraction (XRD), extended X-ray absorption fine structure (EXAFS), and theoretical calculations. In the field of nuclear fuel reprocessing, N,N-dialkyl amides are alternative organic ligands to achieve the separation of uranium(VI) and plutonium(IV) from highly concentrated nitric acid solution. EXAFS analysis combined with XRD shows that the coordination structure of U(VI) is identical in the solution and in the solid state and is independent of the alkyl chain: two amide ligands and four bidentate nitrate ions coordinate the uranyl ion. With linear alkyl chain amides, Pu(IV) also adopt identical structures in the solid state and in solution with two amides and four bidentate nitrate ions. With branched alkyl chain amides, the coordination structure of Pu(IV) was more difficult to establish unambiguously from EXAFS. Density functional theory (DFT) calculations were consequently performed on a series of structures with different coordination modes. Structural parameters and Debye-Waller factors derived from the DFT calculations were used to compute EXAFS spectra without using fitting parameters. By using this methodology, it was possible to show that the branched alkyl chain amides form partly outer-sphere complexes with protonated ligands hydrogen bonded to nitrate ions. PMID:27171842

  15. Standard test method for determination of uranium or plutonium isotopic composition or concentration by the total evaporation method using a thermal ionization mass spectrometer

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This method describes the determination of the isotopic composition and/or the concentration of uranium and plutonium as nitrate solutions by the thermal ionization mass spectrometric (TIMS) total evaporation method. Purified uranium or plutonium nitrate solutions are loaded onto a degassed metal filament and placed in the mass spectrometer. Under computer control, ion currents are generated by heating of the filament(s). The ion beams are continually measured until the sample is exhausted. The measured ion currents are integrated over the course of the run, and normalized to a reference isotope ion current to yield isotopic ratios. 1.2 In principle, the total evaporation method should yield isotopic ratios that do not require mass bias correction. In practice, some samples may require this bias correction. When compared to the conventional TIMS method, the total evaporation method is approximately two times faster, improves precision from two to four fold, and utilizes smaller sample sizes. 1.3 The tot...

  16. Long-range tropospheric transport of uranium and plutonium weapons fallout from Semipalatinsk nuclear test site to Norway.

    Science.gov (United States)

    Wendel, Cato Christian; Fifield, L Keith; Oughton, Deborah H; Lind, Ole Christian; Skipperud, Lindis; Bartnicki, Jerzy; Tims, Stephen G; Høibråten, Steinar; Salbu, Brit

    2013-09-01

    A combination of state-of-the-art isotopic fingerprinting techniques and atmospheric transport modelling using real-time historical meteorological data has been used to demonstrate direct tropospheric transport of radioactive debris from specific nuclear detonations at the Semipalatinsk test site in Kazakhstan to Norway via large areas of Europe. A selection of archived air filters collected at ground level at 9 stations in Norway during the most intensive atmospheric nuclear weapon testing periods (1957-1958 and 1961-1962) has been screened for radioactive particles and analysed with respect to the concentrations and atom ratios of plutonium (Pu) and uranium (U) using accelerator mass spectrometry (AMS). Digital autoradiography screening demonstrated the presence of radioactive particles in the filters. Concentrations of (236)U (0.17-23nBqm(-3)) and (239+240)Pu (1.3-782μBqm(-3)) as well as the atom ratios (240)Pu/(239)Pu (0.0517-0.237) and (236)U/(239)Pu (0.0188-0.7) varied widely indicating several different sources. Filter samples from autumn and winter tended to have lower atom ratios than those sampled in spring and summer, and this likely reflects a tropospheric influence in months with little stratospheric fallout. Very high (236)U, (239+240)Pu and gross beta activity concentrations as well as low (240)Pu/(239)Pu (0.0517-0.077), (241)Pu/(239)Pu (0.00025-0.00062) and (236)U/(239)Pu (0.0188-0.046) atom ratios, characteristic of close-in and tropospheric fallout, were observed in filters collected at all stations in Nov 1962, 7-12days after three low-yield detonations at Semipalatinsk (Kazakhstan). Atmospheric transport modelling (NOAA HYSPLIT_4) using real-time meteorological data confirmed that long range transport of radionuclides, and possibly radioactive particles, from Semipalatinsk to Norway during this period was plausible. The present work shows that direct tropospheric transport of fallout from atmospheric nuclear detonations periodically may have

  17. Uranium

    International Nuclear Information System (INIS)

    The author discusses the contribution made by various energy sources in the production of electricity. Estimates are made of the future nuclear contribution, the future demand for uranium and future sales of Australian uranium. Nuclear power growth in the United States, Japan and Western Europe is discussed. The present status of the six major Australian uranium deposits (Ranger, Jabiluka, Nabarlek, Koongarra, Yeelerrie and Beverley) is given. Australian legislation relevant to the uranium mining industry is also outlined

  18. Uranium

    International Nuclear Information System (INIS)

    The development, prospecting, research, processing and marketing of South Africa's uranium industry and the national policies surrounding this industry form the headlines of this work. The geology of South Africa's uranium occurences and their positions, the processes used in the extraction of South Africa's uranium and the utilisation of uranium for power production as represented by the Koeberg nuclear power station near Cape Town are included in this publication

  19. Development of Remote Plutonium Valence State Analyzer

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>In the Purex process of spent fuel reprocessing, the separation of uranium and plutonium depends on the capability extracting state plutonium ions with various valence by TBP. The separate degree of

  20. Uranium

    International Nuclear Information System (INIS)

    The geological setting of uranium resources in the world can be divided in two basic categories of resources and are defined as reasonably assured resources, estimated additional resources and speculative resources. Tables are given to illustrate these definitions. The increasing world production of uranium despite the cutback in the nuclear industry and the uranium requirements of the future concluded these lecture notes

  1. Properties of the resin Levextrel HDEHP in extraction and chromatography. Application to uranium and some transuranium elements

    International Nuclear Information System (INIS)

    Study of extraction chromatography with a macroporous structure made of a copolymer styrene-divinylbenzene containing di-2-ethylhexylphosphoric acid commercialized under the name Levextrel OC 1026. Uranium, neptunium and plutonium with an oxidation degree of 4 or 6 have a great affinity for Levextrel in perchloric or nitric acid, fixation of neptunium V is low and affinity of americium III varies with acidity this leads to a separation method of actinides. Europium behavior has been specially studied and application to separation of Eu/Nd has been made

  2. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  3. The United States Transuranium and Uranium Registries

    International Nuclear Information System (INIS)

    The United States Transuranium and Uranium Registries are unique parallel research programs devoted to the study of the actinide elements in man. The primary mission of the Registries is to verify and ensure the adequacy and applicability of radiation protection standards for the actinides. To accomplish this task, the Registries utilize tissues obtained postmortem from informed volunteer donors with confirmed or high likelihood of exposure to plutonium, americium, or other actinides. These are collected at autopsy and radiochemically analyzed for actinide content. The results, along with relevant details of occupational and exposure history, medical history and health physics data are used to determine the distribution, biokinetics and dosimetry of the actinides in humans, and to correlate estimates of deposition and dose made during life with postmortem findings. Other important applications of the Registries' research is scaling of animal studies to man and validation or refinement of biokinetic models on which the safety standards are based

  4. SULFIDE METHOD PLUTONIUM SEPARATION

    Science.gov (United States)

    Duffield, R.B.

    1958-08-12

    A process is described for the recovery of plutonium from neutron irradiated uranium solutions. Such a solution is first treated with a soluble sullide, causing precipitation of the plutoniunn and uraniunn values present, along with those impurities which form insoluble sulfides. The precipitate is then treated with a solution of carbonate ions, which will dissolve the uranium and plutonium present while the fission product sulfides remain unaffected. After separation from the residue, this solution may then be treated by any of the usual methods, such as formation of a lanthanum fluoride precipitate, to effect separation of plutoniunn from uranium.

  5. Americium product solidification and disposal

    International Nuclear Information System (INIS)

    The americium product from the TRUEX processing plant needs to be converted into a form suitable for ultimate disposal. An evaluation of the disposal based on safety, number of process steps, demonstrated operability of the processes, production of low-level alpha waste streams, and simplicity of maintenance with low radiation exposures to personnel during maintenance, has been made. The best process is to load the americium on a cation exchange resin followed by calcination or oxidation of the resin after loading

  6. Americium-based oxides: Dense pellet fabrication from co-converted oxalates

    Energy Technology Data Exchange (ETDEWEB)

    Horlait, Denis; Lebreton, Florent [CEA, DEN, DTEC/SDTC/LEMA, 30207 Bagnols-sur-Cèze (France); Gauthé, Aurélie [CEA, DEN, DRCP/SERA/LCAR, 30207 Bagnols-sur-Cèze (France); Caisso, Marie [CEA, DEN, DTEC/SDTC/LEMA, 30207 Bagnols-sur-Cèze (France); Arab-Chapelet, Bénédicte; Picart, Sébastien [CEA, DEN, DRCP/SERA/LCAR, 30207 Bagnols-sur-Cèze (France); Delahaye, Thibaud, E-mail: thibaud.delahaye@cea.fr [CEA, DEN, DTEC/SDTC/LEMA, 30207 Bagnols-sur-Cèze (France)

    2014-01-15

    Mixed oxides are used as nuclear fuels and are notably envisaged for future fuel cycles including plutonium and minor actinide recycling. In this context, processes are being developed for the fabrication of uranium–americium mixed-oxide compounds for transmutation. The purpose of these processes is not only the compliance with fuel specifications in terms of density and homogeneity, but also the simplification of the process for its industrialization as well as lowering dust generation. In this paper, the use of a U{sub 0.85}Am{sub 0.15}O{sub 2±δ} powder synthesized by oxalate co-conversion as a precursor for dense fuel fabrications is assessed. This study notably focuses on sintering, which yielded pellets up to 96% of the theoretical density, taking advantage of the high reactivity and homogeneity of the powder. As-obtained pellets were further characterized to be compared to those obtained via processes based on the UMACS (Uranium Minor Actinide Conventional Sintering) process. This comparison highlights several advantages of co-converted powder as a precursor for simplified processes that generate little dust.

  7. Gestion avancée du Plutonium en REP Complémentarité des cycles thorium et uranium

    OpenAIRE

    Ernoult, Marc

    2014-01-01

    Dans le but d'étudier les possibilités de gestion avancée du plutonium dans les réacteurs actuels, 8 stratégies de multi-recyclage du plutonium dans des REP sont étudiées. Suite à des études à l'équilibre, il a été montré que, en utilisant des assemblages homogènes, l'utilisation du thorium ne permet pas de réduire les inventaires de plutonium en cycle à l'équilibre ni la production d'américium. En répartissant les différents types de combustibles au sein d'un même assemblage, certaines strat...

  8. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 2: A survey of the accuracy of the Studsvik of America CMS codes

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1999-02-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. Thus, before performing any kind of calculation with MOx fuels, it is necessary to be able to establish the reliability and the accuracy of these Core Management System (CMS) codes. This report presents a quantitative analysis of the models used in the package. A qualitative presentation is realized in a coming report.

  9. Room-temperature oxidation of hypostoichiometric uranium-plutonium mixed oxides U1-yPuyO2-x - A depth-selective approach

    Science.gov (United States)

    Vauchy, Romain; Robisson, Anne-Charlotte; Belin, Renaud C.; Martin, Philippe M.; Scheinost, Andreas C.; Hodaj, Fiqiri

    2015-10-01

    In the present work, TGA, XAS and XRD were used to evidence the spontaneous oxidation of biphasic U1-yPuyO2-x samples, with y = 0.28 and 0.45, at room temperature and upon exposure to low moisture and oxygen contents. The oxidation occurs within very short timescales (e.g. O/M ratio increasing from 1.94 to 1.98 within ∼1 μm surface layer in ∼50 h). The combined use of these three complementary methods offered a depth-selective approach from the sample's bulk to its surface and allowed a thorough understanding of the underlying processes involved during the formation of the oxidized layer and of its thickening with time. We believe our results to be of interest in the prospect of fabricating hypo-stoichiometric uranium-plutonium mixed oxides since mastering the oxygen content is a crucial point for many of the fuel properties.

  10. Discrimination of source reactor type by multivariate statistical analysis of uranium and plutonium isotopic concentrations in unknown irradiated nuclear fuel material.

    Science.gov (United States)

    Robel, Martin; Kristo, Michael J

    2008-11-01

    The problem of identifying the provenance of unknown nuclear material in the environment by multivariate statistical analysis of its uranium and/or plutonium isotopic composition is considered. Such material can be introduced into the environment as a result of nuclear accidents, inadvertent processing losses, illegal dumping of waste, or deliberate trafficking in nuclear materials. Various combinations of reactor type and fuel composition were analyzed using Principal Components Analysis (PCA) and Partial Least Squares Discriminant Analysis (PLSDA) of the concentrations of nine U and Pu isotopes in fuel as a function of burnup. Real-world variation in the concentrations of (234)U and (236)U in the fresh (unirradiated) fuel was incorporated. The U and Pu were also analyzed separately, with results that suggest that, even after reprocessing or environmental fractionation, Pu isotopes can be used to determine both the source reactor type and the initial fuel composition with good discrimination.

  11. Pyrochemical conversion of weapon-grade plutonium into plutonium oxide

    International Nuclear Information System (INIS)

    One of the objectives of the French-Russian studies conducted from 1993 to 1996 under the AIDA-MOX 1 program was to define a reference process for converting the weapon-grade plutonium excess (designated W-Pu) into plutonium dioxide for further use as MOX fuel in existing nuclear reactors. Among the different selected options, one is performed in molten alkali chlorides bath at high temperature. Several laboratory-scale tests have permitted to demonstrate the feasibility of this conversion in this medium. The main results described in this paper -conversion yield, plutonium purification beside gallium, americium and other impurities, - tend to confirm that pyrochemical processes could offer potential interests if however the plutonium oxide sinterability is proved in next tests. (authors)

  12. Boiling water reactors with Uranium-Plutonium mixed oxide fuel. Report 1: Accuracy of the nuclide concentrations calculated by CASMO-4

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1999-07-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. These CMS (Core Management System) programs have been extensively compared with both measurements and reference codes. Nevertheless some data are proprietary in particular the comparison of the calculated nuclide concentrations versus experiments (because of the cost of this kind of experimental study). This is why this report describes such a comparative investigation carried out with a General Electric 7x7 BWR bundle. Unfortunately, since some core history parameters were unknown, a lot of hypotheses have been adopted. This invokes sometimes a significant discrepancy in the results without being able to determine the origin of the differences between calculations and experiments. Yet one can assess that, except for four nuclides - Plutonium-238, Curium-243, Curium-244 and Cesium-135 - for which the approximate power history (history effect) can be invoked, the accuracy of the calculated nuclide concentrations is rather good if one takes the numerous approximations into account.

  13. Geochemistry of plutonium and of uranium and thorium series nuclides in the Gulf of Mexico. Summary of progress

    International Nuclear Information System (INIS)

    A summary outline of research on plutonium and naturally occurring radionuclides in the Gulf of Mexico and its contributing rivers is presented. Analytical procedures have been established; river sediments, river water, and marine sediments have been extensively sampled and analyzed for radionuclides and geochemically significant stable elements

  14. Stop plutonium; Stop plutonium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-02-01

    This press document aims to inform the public on the hazards bound to the plutonium exploitation in France and especially the plutonium transport. The first part is a technical presentation of the plutonium and the MOX (Mixed Oxide Fuel). The second part presents the installation of the plutonium industry in France. The third part is devoted to the plutonium convoys safety. The highlight is done on the problem of the leak of ''secret'' of such transports. (A.L.B.)

  15. Technology for Treatment of Liquid Radioactive Waste Generated during Uranium and Plutonium Chemical and Metallurgical Manufacturing in FSUE PO Mayak - 13616

    International Nuclear Information System (INIS)

    Created technological scheme for treatment of liquid radioactive waste generated while uranium and plutonium chemical and metallurgical manufacturing consists of: - Liquid radioactive waste (LRW) purification from radionuclides and its transfer into category of manufacturing waste; - Concentration of suspensions containing alpha-nuclides and their further conversion to safe dry state (calcinate) and moving to long controlled storage. The following technologies are implemented in LRW treatment complex: - Settling and filtering technology for treatment of liquid intermediate-level waste (ILW) with volume about 1500m3/year and alpha-activity from 106 to 108 Bq/dm3 - Membrane and sorption technology for processing of low-level waste (LLW) of radioactive drain waters with volume about 150 000 m3/year and alpha-activity from 103 to 104 Bq/dm3. Settling and filtering technology includes two stages of ILW immobilization accompanied with primary settling of radionuclides on transition metal hydroxides with the following flushing and drying of the pulp generated; secondary deep after settling of radionuclides on transition metal hydroxides with the following solid phase concentration by the method of tangential flow ultrafiltration. Besides, the installation capacity on permeate is not less than 3 m3/h. Concentrates generated are sent to calcination on microwave drying (MW drying) unit. Membrane and sorption technology includes processing of averaged sewage flux by the method of tangential flow ultrafiltration with total capacity of installations on permeate not less than 18 m3/h and sorption extraction of uranium from permeate on anionite. According to radionuclide contamination level purified solution refers to general industrial waste. Concentrates generated during suspension filtering are evaporated in rotary film evaporator (RFE) in order to remove excess water, thereafter they are dried on infrared heating facility. Solid concentrate produced is sent for long

  16. Self-irradiation study of plutonium alloys

    International Nuclear Information System (INIS)

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

  17. PLUTONIUM SEPARATION METHOD

    Science.gov (United States)

    Beaufait, L.J. Jr.; Stevenson, F.R.; Rollefson, G.K.

    1958-11-18

    The recovery of plutonium ions from neutron irradiated uranium can be accomplished by bufferlng an aqueous solutlon of the irradiated materials containing tetravalent plutonium to a pH of 4 to 7, adding sufficient acetate to the solution to complex the uranyl present, adding ferric nitrate to form a colloid of ferric hydroxide, plutonlum, and associated fission products, removing and dissolving the colloid in aqueous nitric acid, oxldizlng the plutonium to the hexavalent state by adding permanganate or dichromate, treating the resultant solution with ferric nitrate to form a colloid of ferric hydroxide and associated fission products, and separating the colloid from the plutonlum left in solution.

  18. Analytical control of reducing agents on uranium/plutonium partitioning at purex process; Controle analitico dos agentes redutores na particao uranio/plutonio no processo purex

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Izilda da Cruz de

    1995-07-01

    Spectrophotometric methods for uranium (IV), hydrazine (N{sub 2}H{sub 4}) and its decomposition product hydrazoic acid(HN{sub 3}), and hydroxylamine (NH{sub 2} 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{sup -6}M for U(IV) with 0,8% of precision, 1,6x10{sup -6}M for hydrazine with 0,8% of precision, 2,3x10{sup -6}M hydrazoic acid with 0,9% of precision and 2,5x10{sup -6}M 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)

  19. On weapons plutonium in the arctic environment (Thule, Greenland)

    International Nuclear Information System (INIS)

    more in mass than 239Pu). This paper concludes that there are at least two different source terms for the Thule accident debris. Another interesting feature is that it seems like uranium is leaching from the particles faster than plutonium and americium. (au)

  20. On weapons plutonium in the arctic environment (Thule, Greenland)

    Energy Technology Data Exchange (ETDEWEB)

    Eriksson, M

    2002-04-01

    (about 4times more in mass than {sup 239}Pu). This paper concludes that there are at least two different source terms for the Thule accident debris. Another interesting feature is that it seems like uranium is leaching from the particles faster than plutonium and americium. (au)

  1. The occurrence of plutonium in nature

    Energy Technology Data Exchange (ETDEWEB)

    Levine, Charles A.; Seaborg, Glenn T.

    1950-11-29

    Plutonium has been chemically separated from seven different ores and the ratios of plutonium to uranium determined. This ratio was found to be fairly constant in pitchblende and monazite ores, in which the uranium content varied from 50% t o 0.24%, and substantially less in carnotite and fergusonite.

  2. Kilogram-scale purification of americium by ion exchange

    Energy Technology Data Exchange (ETDEWEB)

    Wheelwright, E. J.

    1979-01-01

    Sequential anion and cation exchange processes have been used for the final purification of /sup 241/Am recovered during the reprocessing of aged plutonium metallurgical scrap. Plutonium was removed by absorption of Dowex 1, X-3.5 (30 to 50 mesh) anion exchange resin from 6.5 to 7.5 M HNO/sub 3/ feed solution. Following a water dilution to 0.75 to 1.0 M HNO/sub 3/, americium was absorbed on Dowex 50W, X-8 (50 to 100 mesh) cation exchange resion. Final purification was accomplished by elution of the absorbed band down 3 to 4 successive beds of the same resin, preloaded with Zn/sup 2 +/, with an NH/sub 4/OH buffered chelating agent. The recovery of mixed /sup 241/Am-/sup 243/Am from power reactor reprocessing waste has been demonstrated. Solvent extraction was used to recover a HNO/sub 3/ solution of mixed lanthanides and actinides from waste generated by the reprocessng of 13.5 tons of Shippingport Power Reactor blanket fuel. Sequential cation exchange band-displacement processes were then used to separate americium and curium from the lanthanides and then to separate approx. 60 g of /sup 244/Cm from 1000 g of mixed /sup 241/Am-/sup 243/Am.

  3. Analytical performance of radiochemical method for americium determination in urine

    International Nuclear Information System (INIS)

    This paper presents an analytical method developed and adapted for separation and analysis of Plutonium (Pu) isotopes and Americium (Am) in urine samples. The proposed method will attend the demand of internal exposure monitoring program for workers involved mainly with dismantling rods and radioactive smoke detectors. In this experimental procedure four steps are involved as preparation of samples, sequential radiochemical separation, preparation of the source for electroplating and quantification by alpha spectrometry. In the first stage of radiochemical separation, plutonium is conventionally isolated employing the anion exchange technique. Americium isolation is achieved sequentially by chromatographic extraction (Tru.spec column) from the load and rinse solutions coming from the anion exchange column. The 243Am tracer is added into the sample as chemical yield monitors and to correct the results improving the precision and accuracy. The mean recovery obtained is 60%, and the detection limit for 24h urine sample is 1.0 mBq L-1 in accordance with the literature. Based in the preliminary results, the method is appropriate to be used in monitoring programme of workers with a potential risk of internal contamination. (author)

  4. Determination of plutonium isotopes in samples of the Chernobyl exclusion zone by means of characteristic Ix-radiation of uranium

    International Nuclear Information System (INIS)

    The method for measurement of alpha-emitting Pu isotopes by means of characteristic Lx-radiation of uranium in environmental samples has been proposed. This method was compared with standard radiochemistry determination of Pu isotopes in soil samples from the Red forest. The results of comparison between both methods have a high correlation

  5. Determination of Isotopic Composition of Uranium and Plutonium in Simulated 1AF Feed Solution by TIMS%TIMS测定模拟1AF料液中铀钚同位素组分

    Institute of Scientific and Technical Information of China (English)

    蒋军清; 高月华; 汪南杰; 姜国杜

    2012-01-01

    The precise determination of the uranium and plutonium isotopic composition in the simulated 1AF feed solution by solvent extraction/ion exchange separation-thermal ioni-zation mass spectroscopy(TIMS) was studied in this paper. The precise determination of uranium and plutonium isotopic composition was realized by optimization of chemical separations, and minimization of possible interferences, and proper selection of instrument parameters. Under the selected conditions, the uranium and plutonium isotopic composition in the 1AF feed solution were accurately determined. The relative standard deviations of the main isotopes 235U and 239Pu are less than 0. 05%.%采用溶剂萃取/离子交换分离-热表面电离质谱法,对模拟1AF料液中铀钚同位素组分测定技术进行了研究.通过对化学分离条件、仪器测量参数、信号强度、各种干扰等测定条件的研究和选择,实现了铀、钚同位素组分的精密测定.在选定的条件下,测定了模拟1AF料液中的铀钚同位素,主要同位素235 U和239pu测定精密度(sr)均优于0.05%.

  6. Approach to derive doses for case-control studies of lung cancer and leukaemia among workers internally exposed to uranium and plutonium

    International Nuclear Information System (INIS)

    Case-control studies are currently conducted in 3 European countries (Belgium, France and the United Kingdom) to estimate the risk of lung cancer and leukaemia in relation to internal exposure to uranium and plutonium amongst workers in the nuclear industry. The project requires calculating doses absorbed by the lung and the bone marrow for many hundreds of cases and controls internally exposed. In order to establish a common approach to dose reconstruction, a detailed dosimetry protocol and a database of individual exposure were set up and will be presented. The dose reconstruction relies heavily on bioassay data, which are usually urine analysis, extending back over 50 years in some cases. Inevitably, data obtained over such a time span are of variable quality. It is important to review the monitoring practices at the various laboratories and to assess the reliability of these data in order to estimate possible biases as well as random uncertainties. Another key step in the reconstruction process is to decide upon the likely intake regimes consistent with the data. Generally, chronic intakes will be assumed and acute intakes will be added only when their existence is supported by operational data. Biokinetic models are used both to calculate intakes from bioassay data and to convert intakes to doses. The ICRP publication 66 respiratory tract model will be used along with the latest systemic models described by ICRP. These will be supplemented by the Leggett 2005 model for plutonium. These various models will be implemented by the code IMBA-Expert. Since it is essential to obtain central estimates for the doses, a particular problem is encountered with datasets consisting only of values below the limit of detection. For these cases Bayesian statistics will be employed using a non-informative prior probability distribution. (author)

  7. Reactivity of the uranium (U(IV)/U(VI)) and the plutonium (Pu(III)/Pu(IV)) in nitric aqueous solution under ultrasound

    International Nuclear Information System (INIS)

    To minimize the volumes of solid waste and industrial effluents generated at the end of cycle, particularly in the spent nuclear fuel reprocessing industry, research is currently under way on so-called innovative processes, designed to induce chemical reactions without adding reagent to the media. Among these processes, the use of ultrasound can prove advantageous, and the purpose of this study is to assess accurately the potential for its application. In the present context, this work shows that the transmission of an ultrasonic wave in aqueous nitric acid solution leads to: the accumulation of nitrous acid in solution, until a steady-sate concentration is reached; the removal of nitrogen monoxide and nitrogen dioxide in the gas stream. The initial kinetics of the formation of HNO2 in solution was quantified as a function of the nitric acid concentration and the ultrasound intensity. It was also shown than an excess of nitrous acid in nitric solution decomposes under the effect of ultrasound. It is also possible to accumulate hydrogen peroxide in solution during the ultrasonic irradiation of aqueous nitric acid solutions in the presence of a chemical species N2H5+, NH2SO3H...) which reacts rapidly with HNO2, preventing the reduction of H2O2 by HNO2. The mechanisms of HNO2 formation and decomposition, and the mechanism of H2O2 formation during the ultrasonic irradiation of aqueous nitric acid solutions, are presented. Control of H2O2 or HNO2 in a nitric acid medium under the effect of an ultrasonic wave can be exploited to control redox reactions of uranium and plutonium ions, particularly with respect to the oxidation of U and Pu (U(IV)→ U(IV) or Pu(III) → Pu(IV)) and the reduction of Pu (Pu(IV)→ Pu(III). The redox behavior of uranium and plutonium ions in aqueous nitric solution subject to an ultrasonic flux is interpreted in term of effects induced on the reaction medium, and reveals the potential for using ultrasound to cause chemical reactions in

  8. Phosphate monazite- and NaZr2(PO4)3 (NZP)-like ceramics containing uranium and plutonium

    Science.gov (United States)

    Orlova, A. I.; Kitaev, D. B.; Lukinich, A. N.; Tomilin, S. B.; Lizin, A. A.; Kulikov, I. A.; Samsonov, V. E.

    2003-01-01

    The synthesis and results of the X-ray diffraction examination of complex orthophosphates involving the cations Ca2+, Cd2+, Gd3+, Ti4+, Hf4+, Ce4+, U4+, and Pu4+ are described. Crystalline products containing either cerium or plutonium or these two elements together were prepared. Cerium was used as plutonium simulator. Phosphates containing cations Ti4+ and Hf4+ were found to form two-phase products. One of these phases was assumed to be of the NaZr2(PO4)3 (NZP) type, the other one of the CePO4 type (monazite). Formation of single-phase products with a monazite structure was found in the specimens that did not contain Ti and Hf. X-ray diffraction characteristics of synthesised phases were identified. An effect of the cation composition on the phase formation and also the effect of the cation size on the crystal lattice parameters of the formed phases were established.

  9. Rapid assay of plutonium in soils by passive L x-ray counting

    Energy Technology Data Exchange (ETDEWEB)

    Gehrke, R.J.; Putnam, M.H.; Goodwin, S.G.; Kynaston, R.L.

    1992-05-01

    A technique has been developed to rapidly measure the presence of plutonium in soils, filters, smears, and glass waste forms by measuring the uranium L-shell x-ray emissions associated with the decay of plutonium. In addition, the technique can simultaneously acquire spectra of samples and automatically analyze them for the amount of americium, and gamma-ray emitting activation and fission products present. The samples are counted with a large area, thin-window, n-type Ge spectrometer which is equally efficient for the detection of low energy x-rays (>10 key), as well as high-energy gamma rays (>1 MeV). A 8192-channel analyzer is used to acquire the entire photon spectrum at one time. A dual-energy, time-tagged pulser, that is injected into the test input of the preamplifier to monitor the energy scale, detector resolution, and pulse pile-up will be installed in FY-92. The L x-ray portion of each spectrum is analyzed by a linear least-squares spectral fitting technique originally developed for the analysis of spectra from NaI(Tl) detectors. The gamma-ray portion of each spectrum is analyzed by a standard Ge gamma-ray analysis package. Detection limits (also referred to as lower limits of detection) for plutonium in contaminated soils that have been achieved by this technique are reported.

  10. Higher Americium Oxidation State Research Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Mincher, Bruce J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Law, Jack D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Goff, George S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Moyer, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burns, Jon D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lumetta, Gregg J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sinkov, Sergey I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Shehee, Thomas C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hobbs, David T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-12-18

    -solvent-extraction separations are also under investigation. The first would separate Am(VI) by co-crystallization with uranium and the other oxidizable actinides as their nitrate salts. This novel idea has been successful in lab scale testing, and merits further investigation. Similarly, success has been achieved in separations using inorganic or hybrid ion exchange materials to sorb the lanthanides and actinides, while allowing pentavalent americium to elute. This is the only technique currently investigating Am(V), despite the advantages of this oxidation state with regard to its higher stability. The ultimate destination for this roadmap is to develop an americium separation that can be applied under process conditions, preferably affording a co-separation of the actinyl (VI) ions. Toward that end, emphasis is given here to selection of a solvent extraction flowsheet for testing in the INL centrifugal contactor hot test bed during FY16. A solvent extraction process will be tested mainly because solvent extraction separations of Am(VI) are relatively mature and the test bed currently exists in a configuration to support them. Thus, a major goal of FY16 is to select the oxidant/ligand combination to run such a test using the contactors. The only ligands under consideration are DAAP and DEHBA. This is not to say that ion exchange and co-crystallization techniques are unimportant. They merit continued investigation, but are not mature enough for hot test bed testing at this time.

  11. Synthesis and extraction studies with a rationally designed diamide ligand selective to actinide(iv) pertinent to the plutonium uranium redox extraction process.

    Science.gov (United States)

    Sharma, Shikha; Panja, Surajit; Bhattacharyya, Arunasis; Dhami, Prem S; Gandhi, Preetam M; Ghosh, Sunil K

    2016-05-01

    A new class of conformationally constrained oxa-bridged tricyclo-dicarboxamide (OTDA) ligand was rationally designed for the selective extraction of tetravalent actinides pertinent to the Plutonium Uranium Redox EXtraction (PUREX) process. Two of the designed diamide ligands were synthesized and extraction studies were performed for Pu(iv) from HNO3 medium. The mechanism of extraction was investigated by studying various parameters such as feed HNO3, NaNO3 and OTDA concentrations. The nature of the extracted species was found to be [Pu(NO3)4(OTDA)]. One of the OTDA ligands was elaborately tested and showed the selective extraction of Pu(iv) and Np(iv) over other actinide species, viz., U(vi), Np(v), Am(iii), lanthanides and fission products contained in a nuclear waste from the PUREX process. DFT calculations predicted the charge density on each of the coordinating 'O' atoms of OTDA supporting its high Pu(iv) selectivity over other ions studied and also provided the energy optimized structure of OTDA and its Pu(iv) complex. PMID:27054892

  12. The effects of fuel microstructure evolution on thermal performance in fast reactor type uranium-plutonium oxide fuel pins up to high burnup

    International Nuclear Information System (INIS)

    Higher power density and larger thermal gradients in fast reactor type uranium-plutonium oxide fuel pins than in water reactor type fuel rods induces the drastic fuel microstructure evolutions, which are useful for estimating the thermal condition at power in the fuel pellets. Recently, fuel-to-cladding gap reopen phenomena is paid attention to due to the influences on the heat transfer behaviors. The objectives of this work are to clarify the dominant factors for fuel microstructure evolution and investigate the gap reopen phenomena on the fuel pin thermal performance up to high burnup. The results of irradiation tests accumulated in JOYO Mk-II core, FFTF, Phenix were exhaustively collected and compiled to the database in this work. Then, the influences of irradiation conditions on the extent of microstructure evolution, such as central void, columnar grain, gas bubble region, densified region, dark ring, and residual gap (including gap reopen phenomena), were reviewed and investigated. Also the thermal behaviors in fuel pins were modeled to apply for the thermal analysis based on the recent knowledge referred to water reactor fuel experiences. It was revealed that the gap reopen phenomena related directly with the fission gas release behaviors in JOYO Mk-II J2 type driver fuel pins whose fabrication, irradiation, and post-irradiation conditions were well characterized. The results of one dimensional thermal analysis at the transverse sections were proved that gap conductance were improved by the condensed material which fills the gap even after the gap reopen phenomena. (author)

  13. A method for determining an effective porosity correction factor for thermal conductivity in fast reactor uranium-plutonium oxide fuel pellets

    International Nuclear Information System (INIS)

    A reliable method has been developed for determining an effective porosity correction factor for calculating a realistic thermal conductivity for fast reactor uranium-plutonium (mixed) oxide fuel pellets. By using image analysis of the ceramographs of transverse sections of mixed-oxide fuel pellets, the fuel morphology could be classified into two basic types. One is a 'two-phase' type that consists of small pores dispersed in the fuel matrix. The other is a 'three-phase' type that has large pores in addition to the small pores dispersed in the fuel matrix. The pore sizes are divided into two categories, large and small, at the 30 μm area equivalent diameter. These classifications lead to an equation for calculating an effective porosity correction factor by accounting for the small and large pore volume fractions and coefficients. This new analytical method for determining the effective porosity correction factor for calculating the realistic thermal conductivity of mixed-oxide fuel was also experimentally confirmed for high-, medium- and low-density fuel pellets

  14. Extraction behavior of uranium(VI), plutonium(VI) and some fission products with gamma pre-irradiated n-dodecane solutions of n,n'-dihexyl substituted amides

    International Nuclear Information System (INIS)

    The extraction behavior of uranium(VI), plutonium(IV) and some fission products like zirconium(IV), ruthenium(III) and europium(III) from 3.5M nitric acid with γ-irradiated organic phase pre-equilibrated n-dodecane solutions of dihexyl derivatives of hexanamide (DHHA), octanamide (DHOA) and decanamide (DHDA) has been investigated as a function of absorbed dose upto 184x104 Gy. The results indicate that the extraction of uranium(VI) decreases gradually with dose upto 72x104 Gy and becomes almost constant thereafter, while, the extraction of plutonium(IV) decreases upto a dose of 20x104 Gy and then increases rapidly up to a dose of 82x104 Gy indicating synergistic effects of radiolytic products formed at higher doses. Extraction of zirconium(IV) increases gradually upto a dose 72x104 Gy. Europium(III) does not get extracted with any of these amides in the entire dose range (0-184x104 Gy) studied, however, ruthenium shows insignificant increase in extraction with dose. The decrease in D values noticed in the case of plutonium and zirconium after the dose of 72x104 Gy which was attributed to the third phase formation and emulsification. Infrared studies confirm the final products of radiolysis as the respective amines and carboxylic acids. The degraded amide contents have been estimated by quantitative IR spectrophotometric technique. Extraction data obtained for uranium(VI) and plutonium(IV) with TBP/n-dodecane system have also been compared under similar experimental conditions. (author). 30 refs., 1 fig., 3 tabs

  15. Impact of the cation distribution homogeneity on the americium oxidation state in the U0.54Pu0.45Am0.01O2-x mixed oxide

    Science.gov (United States)

    Vauchy, Romain; Robisson, Anne-Charlotte; Martin, Philippe M.; Belin, Renaud C.; Aufore, Laurence; Scheinost, Andreas C.; Hodaj, Fiqiri

    2015-01-01

    The impact of the cation distribution homogeneity of the U0.54Pu0.45Am0.01O2-x mixed oxide on the americium oxidation state was studied by coupling X-ray diffraction (XRD), electron probe micro analysis (EPMA) and X-ray absorption spectroscopy (XAS). Oxygen-hypostoichiometric Am-bearing uranium-plutonium mixed oxide pellets were fabricated by two different co-milling based processes in order to obtain different cation distribution homogeneities. The americium was generated from β- decay of 241Pu. The XRD analysis of the obtained compounds did not reveal any structural difference between the samples. EPMA, however, revealed a high homogeneity in the cation distribution for one sample, and substantial heterogeneity of the U-Pu (so Am) distribution for the other. The difference in cation distribution was linked to a difference in Am chemistry as investigated by XAS, with Am being present at mixed +III/+IV oxidation state in the heterogeneous compound, whereas only Am(IV) was observed in the homogeneous compound. Previously reported discrepancies on Am oxidation states can hence be explained by cation distribution homogeneity effects.

  16. Surface Reduction of Neptunium Dioxide and Uranium Mixed Oxides with Plutonium and Thorium by Photocatalytic Reaction with Ice

    OpenAIRE

    CAKIR PELIN; ELOIRDI Rachel; Huber, Frank; KONINGS Rudy; GOUDER Thomas

    2014-01-01

    The surface reductions of neptunium dioxide (NpO2) and two mixed oxides of uranium (U−Pu−O2 and U−Th−O2) with adsorbed water ice were studied by ultraviolet and X-ray photoelectron spectroscopy (UPS and XPS, respectively). The oxides were produced as thin films by reactive sputter deposition. Water was condensed as a thick ice overlayer on the surface at low temperature. Subsequent warming led to desorption of the ice. When warmed up under ultraviolet light (UV light, HeI and HeII radiatio...

  17. On complexe oxalates of plutonium(3)

    International Nuclear Information System (INIS)

    Plutonium compounds of general composition MPu(C2O4)2xnH2O, where M=Li, Na, NH4, K, Cs and n=0.5-3.5, have been separated from solutions of alkali metal and ammonium oxalates and identified using the methods of X-ray diffraction and chemical analyses. The behaviour of plutonium (3) differs from that of americium (3) in solutions of lithium- and sodium oxalates, but the difference decreases with the increase in the monovalent cation radius, so that in the cesium oxalate solution plutonium (3) behaves in the same manner as americium (3). The reaction of the compounds NH4Pu(C2O4)2xH2O and CsPu(C2O4)2x0.5H2O to heating is studied. Ammonium salt of plutonium oxalate begins to rapidly decompose at 250 deg C. At approximately 350 deg C the formation of PuO2 completes. Rapid decomposition of cesium salt of plutonium oxalate is observed above 300 deg C. The study of the IR spectra of double plutonium (3) with ammonium-or-cesium oxalates permitted to draw certain conclusions on the nature of chemical bonds in the compounds

  18. Qualitative chemical analysis of plutonium by Alpha spectroscopy

    International Nuclear Information System (INIS)

    In this work the separation and purification of plutonium from irradiated uranium was done. The plutonium, produced by the irradiation of uranium in a nuclear reactor and the β decay of 239 Np, was stabilized to Pu +4 with sodium nitrite. Plutonium was separated from the fission products and uranium by ion exchange using the resin Ag 1 X 8. It was electrodeposited on stainless steel discs and the alpha radioactivity of plutonium was measured in a surface barrier detector. The results showed that plutonium was separated with a radiochemical purity higher than 99 %. (Author)

  19. Uranium Recovery in LWR Reprocessing and Plutonium/Residual Uranium Conditioning in FBR Reprocessing for the Transition from LWR to FBR

    Energy Technology Data Exchange (ETDEWEB)

    Tetsuo Fukasawa; Junichi Yamashita; Kuniyoshi Hoshino [Hitachi-GE Nuclear Energy, Ltd., 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Akira Sasahira [Hitachi, Ltd. (Japan); Tadashi Inoue [Central Research Institute of Electric Power Industry (Japan); Kazuo Minato [Japan Atomic Energy Agency (Japan); Seichi Sato [Hokkaido University (Japan)

    2008-07-01

    In order to flexibly manage the transition period from LWR (light water reactor) to FBR (fast breeder reactor), the authors investigated the transition scenario and proposed the Flexible Fuel Cycle Initiative (FFCI). In FFCI, LWR spent fuel reprocessing only carries out the removal of about 90% uranium that will be purified and utilized in LWR after re-enrichment. The residual material ({approx}40% U, {approx}15% Pu and {approx}45% other nuclides) is transferred to temporary storage and/or FBR spent fuel reprocessing to recover Pu/U followed by FBR fresh fuel fabrication depending on the FBR introduction status. The FFCI has some merits compared with ordinary system that consists of full reprocessing facilities for both LWR and FBR spent fuels, that is smaller LWR reprocessing facility, spent LWR fuel reduction, storage and supply of high proliferation resistant and high Pu density material that can flexibly respond to FBR introduction rate changes. The Pu balance was calculated under several cases, which revealed that the FFCI could supply enough Pu to FBR in any cases. (authors)

  20. Stop plutonium

    International Nuclear Information System (INIS)

    This press document aims to inform the public on the hazards bound to the plutonium exploitation in France and especially the plutonium transport. The first part is a technical presentation of the plutonium and the MOX (Mixed Oxide Fuel). The second part presents the installation of the plutonium industry in France. The third part is devoted to the plutonium convoys safety. The highlight is done on the problem of the leak of ''secret'' of such transports. (A.L.B.)

  1. Increase in oxygen-to-metal (O/M) ratio of sintered pellet of uranium-plutonium mixed oxide stored at room-temperature

    International Nuclear Information System (INIS)

    A number of investigations have been reported so far to reduce the O/M ratio of the sintered pellets of uranium-plutonium mixed oxide for the prevention of fuel-cladding chemical interaction during irradiation. However, little attention has been paid to the phenomenon that the O/M ratio gradually increases in usual glove-box atmosphere. This phenomenon is due to room-temperature oxidation reaction, and is generally accelerated by the humidity in that atmosphere. In order to investigate this phenomenon, periodic O/M measurements were carried out on the pellets stored in three types of atmosphere, 2.4N HCl (wet box), air, and silica gel application (dry box). The pellet specimens used were made of hypostoichiometric mixed oxide, fabricated as FBR type fuel pellets. According to the continuous measurements of the weight of specimens, the O/M increment was estimated for the total storage time up to 130 days. The O/M ratio was determined by heating the specimens to 800 deg C, exposing to air flow for 0.5 hr, and then to N2-5% H2 gas flow for 4.0 hr. It is seen from the experimental curves that the rate of O/M increment apparently depends on the moisture content of running atmosphere. Using the data obtained, the equation to determine the O/M ratio at storage time t was derived, and it is very instructive for presuming the increment of O/M ratio of sintered pellets during storage. (Wakatsuki, Y.)

  2. A quantitative Monte Carlo modelling of the uranium and plutonium X-ray fluorescence (XRF) response from a hybrid K-edge/K-XRF densitometer

    Energy Technology Data Exchange (ETDEWEB)

    Berlizov, A.N., E-mail: berlizov@kinr.kiev.u [National Academy of Sciences of Ukraine, Institute for Nuclear Research, 47, Prospekt Nauki, MSP 03680, Kiev (Ukraine); Institute for Transuranium Elements, Joint Research Centre, European Commission, P.O. Box 2340, D-76125 Karlsruhe (Germany); Sharikov, D.A., E-mail: sharikov@tpu.r [Institute for Transuranium Elements, Joint Research Centre, European Commission, P.O. Box 2340, D-76125 Karlsruhe (Germany); Tomsk Polytechnic University, 30, Lenin Avenue, 634050, Tomsk (Russian Federation); Ottmar, H., E-mail: herbert.ottmar@t-online.d [Institute for Transuranium Elements, Joint Research Centre, European Commission, P.O. Box 2340, D-76125 Karlsruhe (Germany); Eberle, H., E-mail: heinrich.eberle@ec.europa.e [Institute for Transuranium Elements, Joint Research Centre, European Commission, P.O. Box 2340, D-76125 Karlsruhe (Germany); Galy, J., E-mail: jean.galy@ec.europa.e [Institute for Transuranium Elements, Joint Research Centre, European Commission, P.O. Box 2340, D-76125 Karlsruhe (Germany); Luetzenkirchen, K., E-mail: klaus-richard.luetzenkirchen@ec.europa.e [Institute for Transuranium Elements, Joint Research Centre, European Commission, P.O. Box 2340, D-76125 Karlsruhe (Germany)

    2010-03-21

    A mathematical simulation approach based on the general purpose Monte Carlo N-particle transport code MCNP was developed to predict the response of the XRF branch of the hybrid K-edge/K-XRF densitometer (HKED). The respective MCNP models for two different versions of HKED instruments currently in use were set up and experimentally validated. The setting up of the models involved comprehensive simulations of a bremsstrahlung photon source, the examination of different particle transport models, as well as the examination of different photon attenuation and X-ray fluorescence data libraries. The computation speed was significantly increased through the extensive use of the variance reduction techniques. The models were validated through the series of benchmarking experiments performed with a representative set of uranium, plutonium and mixed U/Pu reference solutions. The models and simulation approach developed are intended for: (i) establishing a consistent mathematical calibration approach for the XRF branch of the HKED instruments, which will require minimum calibration effort and time, (ii) extending the applicability of the HKED method to non-standard samples (e.g. U/Pu mixtures with unusual element ratios) and non-standard sample matrices (e.g. HM matrices from the pyro-processing of irradiated nuclear fuel) without investing a great deal of extra calibration work, and (iii) improving the accuracy of the measurements through the modelling of special measurement effects (e.g. the secondary excitation effect, the interference with X-ray escape peaks, the inconsistent unfolding of the overlapping peaks and peak background delineation in the measured XRF spectrum), which are difficult or sometimes impossible to account for experimentally.

  3. Airborne Plutonium and non-natural Uranium from the Fukushima DNPP found at 120 km distance a few days after reactor hydrogen explosions.

    Science.gov (United States)

    Shinonaga, Taeko; Steier, Peter; Lagos, Markus; Ohkura, Takehisa

    2014-04-01

    Plutonium (Pu) and non-natural uranium (U) originating from the Fukushima Daiichi Nuclear Power Plant (FDNPP) were identified in the atmosphere at 120 km distance from the FDNPP analyzing the ratio of number of atoms, following written as n(isotope)/n(isotope), of Pu and U. The n((240)Pu)/n((239)Pu), n((241)Pu)/n((239)Pu), n((234)U)/n((238)U), n((235)U)/n((238)U) and n((236)U)/n((238)U) in aerosol samples collected before and after the FDNPP incident were analyzed by accelerator mass spectrometry (AMS) and inductively coupled plasma mass spectrometry (ICPMS). The activity concentrations of (137)Cs and (134)Cs in the same samples were also analyzed by gamma spectrometry before the destructive analysis. Comparing the time series of analytical data on Pu and U obtained in this study with previously reported data on Pu, U, and radioactive Cs, we concluded that Pu and non-natural U from the FDNPP were transported in the atmosphere directly over a 120 km distance by aerosol and wind within a few days after the reactor hydrogen explosions. Effective dose of Pu were calculated using the data of Pu: (130 ± 21) nBq/m(3), obtained in this study. We found that the airborne Pu contributes only negligibly to the total dose at the time of the incident. However the analytical results show that the amount of Pu and non-natural U certainly increased in the environment after the incident.

  4. Quantum chemistry study of uranium(VI), neptunium(V), and plutonium(IV,VI) complexes with preorganized tetradentate phenanthrolineamide ligands.

    Science.gov (United States)

    Xiao, Cheng-Liang; Wu, Qun-Yan; Wang, Cong-Zhi; Zhao, Yu-Liang; Chai, Zhi-Fang; Shi, Wei-Qun

    2014-10-20

    The preorganized tetradentate 2,9-diamido-1,10-phenanthroline ligand with hard-soft donors combined in the same molecule has been found to possess high selectivity toward actinides in an acidic aqueous solution. In this work, density functional theory (DFT) coupled with the quasi-relativistic small-core pseudopotential method was used to investigate the structures, bonding nature, and thermodynamic behavior of uranium(VI), neptunium(V), and plutonium(IV,VI) with phenanthrolineamides. Theoretical optimization shows that Et-Tol-DAPhen and Et-Et-DAPhen ligands are both coordinated with actinides in a tetradentate chelating mode through two N donors of the phenanthroline moiety and two O donors of the amide moieties. It is found that [AnO2L(NO3)](n+) (An = U(VI), Np(V), Pu(VI); n = 0, 1) and PuL(NO3)4 are the main 1:1 complexes. With respect to 1:2 complexes, the reaction [Pu(H2O)9](4+)(aq) + 2L(org) + 2NO3(-)(aq) → [PuL2(NO3)2](2+)(org) + 9H2O(aq) might be another probable extraction mechanism for Pu(IV). From the viewpoint of energy, the phenanthrolineamides extract actinides in the order of Pu(IV) > U(VI) > Pu(VI) > Np(V), which agrees well with the experimental results. Additionally, all of the thermodynamic reactions are more energetically favorable for the Et-Tol-DAPhen ligand than the Et-Et-DAPhen ligand, indicating that substitution of one ethyl group with one tolyl group can enhance the complexation abilities toward actinide cations (anomalous aryl strengthening).

  5. Recovery of neptunium, plutonium, and americium from highly active waste

    International Nuclear Information System (INIS)

    Trialkylphosphine oxides (TRPO) (alkyl is 6c-C8) were chosen as the extractant for the recovery of Np, Pu, and Am from highly active waste (HAW) because of its extraction ability, excellent solvent behavior, high radiolytic stability, and low cost. Process chemistry based on 30 vol % TRPO-kerosene as solvent is presented. Extraction of Am in the presence of macro amounts of neodymium, adjustment of Np valence by electrolytic reduction, selective stripping of actinides from loaded organic phase, and loading capacity of the solvent are included. Process parameters of multistage countercurrent extraction and stripping and the results of experimental verification are given. From HAW with ∼1 M nitric acid concentration, recovery of actinides is higher than 99.9%. The actinides extracted can be stripped out separately into Am, Np-Pu, and U fractions. The behavior of nonactinide HAW constituents, including Tc, is discussed

  6. Weapons-grade plutonium dispositioning. Volume 4

    International Nuclear Information System (INIS)

    This study is in response to a request by the Reactor Panel Subcommittee of the National Academy of Sciences (NAS) Committee on International Security and Arms Control (CISAC) to evaluate the feasibility of using plutonium fuels (without uranium) for disposal in existing conventional or advanced light water reactor (LWR) designs and in low temperature/pressure LWR designs that might be developed for plutonium disposal. Three plutonium-based fuel forms (oxides, aluminum metallics, and carbides) are evaluated for neutronic performance, fabrication technology, and material and compatibility issues. For the carbides, only the fabrication technologies are addressed. Viable plutonium oxide fuels for conventional or advanced LWRs include plutonium-zirconium-calcium oxide (PuO2-ZrO2-CaO) with the addition of thorium oxide (ThO2) or a burnable poison such as erbium oxide (Er2O3) or europium oxide (Eu2O3) to achieve acceptable neutronic performance. Thorium will breed fissile uranium that may be unacceptable from a proliferation standpoint. Fabrication of uranium and mixed uranium-plutonium oxide fuels is well established; however, fabrication of plutonium-based oxide fuels will require further development. Viable aluminum-plutonium metallic fuels for a low temperature/pressure LWR include plutonium aluminide in an aluminum matrix (PuAl4-Al) with the addition of a burnable poison such as erbium (Er) or europium (Eu). Fabrication of low-enriched plutonium in aluminum-plutonium metallic fuel rods was initially established 30 years ago and will require development to recapture and adapt the technology to meet current environmental and safety regulations. Fabrication of high-enriched uranium plate fuel by the picture-frame process is a well established process, but the use of plutonium would require the process to be upgraded in the United States to conform with current regulations and minimize the waste streams

  7. 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 Director General has received a Note Verbale dated 3 August 2005 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 2004

  8. 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 Director General has received a Note Verbale dated 29 August 2006 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 2005

  9. 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 26 June 2009 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 2008

  10. 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 23 August 2007 from the Permanent Mission of the United Kingdom of Great Britain and Northern Ireland to the United Nations and the International Organizations in Vienna 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 2006

  11. 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 28 July 2010 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 2009

  12. 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 17 June 2013 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 2012

  13. 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 3 July 2012 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 2011

  14. 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 2 August 2004 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 2003

  15. Communication Received from the United Kingdom of Great Britain and Northern Ireland Concerning Its Policies Regarding the Management of Plutonium. Statement on the Management of Plutonium and of High Enriched Uranium

    International Nuclear Information System (INIS)

    The Secretariat has received a Note Verbale dated 17 June 2008 from the Permanent Mission of the United Kingdom of Great Britain and Northern Ireland to the United Nations and the International Organizations in Vienna 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 2007

  16. Reactivity of the uranium (U(IV)/U(VI)) and the plutonium (Pu(III)/Pu(IV)) in nitric aqueous solution under ultrasound; De l'influence des ultrasons sur la reactivite de l'uranium (U(IV)/U(VI)) et du plutonium (PU(III)/PU(IV)) en solution aqueuse nitrique

    Energy Technology Data Exchange (ETDEWEB)

    Venault, L

    1998-07-01

    To minimize the volumes of solid waste and industrial effluents generated at the end of cycle, particularly in the spent nuclear fuel reprocessing industry, research is currently under way on so-called innovative processes, designed to induce chemical reactions without adding reagent to the media. Among these processes, the use of ultrasound can prove advantageous, and the purpose of this study is to assess accurately the potential for its application. In the present context, this work shows that the transmission of an ultrasonic wave in aqueous nitric acid solution leads to: the accumulation of nitrous acid in solution, until a steady-sate concentration is reached; the removal of nitrogen monoxide and nitrogen dioxide in the gas stream. The initial kinetics of the formation of HNO{sub 2} in solution was quantified as a function of the nitric acid concentration and the ultrasound intensity. It was also shown than an excess of nitrous acid in nitric solution decomposes under the effect of ultrasound. It is also possible to accumulate hydrogen peroxide in solution during the ultrasonic irradiation of aqueous nitric acid solutions in the presence of a chemical species (N{sub 2}H{sub 5}{sup +}, NH{sub 2}SO{sub 3}H...) which reacts rapidly with HNO{sub 2}, preventing the reduction of H{sub 2}O{sub 2} by HNO{sub 2}. The mechanisms of HNO{sub 2} formation and decomposition, and the mechanism of H{sub 2}O{sub 2} formation during the ultrasonic irradiation of aqueous nitric acid solutions, are presented. Control of H{sub 2}O{sub 2} or HNO{sub 2} in a nitric acid medium under the effect of an ultrasonic wave can be exploited to control redox reactions of uranium and plutonium ions, particularly with respect to the oxidation of U and Pu (U(IV){yields} U(IV) or Pu(III) {yields} Pu(IV)) and the reduction of Pu (Pu(IV){yields} Pu(III). The redox behavior of uranium and plutonium ions in aqueous nitric solution subject to an ultrasonic flux is interpreted in term of effects

  17. Chrono-potentiometry in molten chlorides. Application to the study of the electrochemical properties of uranium and plutonium in the LiCl-KCl eutectic; Chronopotentiometrie dans les chlorures fondus. Application a l'etude des proprietes electrochimiques de l'uranium et du plutonium dans l'eutectique LiCl-KCl

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-05-01

    Using solutions of cadmium chloride in the eutectic LiCl-KCl, a chrono-potentiometric method has been developed with a view to its application to the study of molten solutions. Particular attention has been paid to the choice of the indicator electrodes. The method makes it possible to analyze molten solutions quantitatively and to determine diffusion coefficients and their activation energies; it yields furthermore information about the nature and the behaviour of ionic species in solution. The method has been applied to the study of solutions of uranium and plutonium chloride in the eutectic LiCl-KCl. Linear chrono-amperometry has been used for studying these solutions quantitatively, but chrono-potentiometry, of which the theory is better developed, is better suited to a quantitative study. The results obtained have made it possible to determine the diffusion coefficients of the ions Cd{sup 2+}, U{sup 3+}, U{sup 4+} and U(IV) in the presence of F{sup -} and Pu{sup 3+} ions, as well as the activation energy of the diffusion coefficients. (author) [French] La mise au point de la chronopotentiometrie comme moyen d'etude des solutions fondues a ete effectuee avec des solutions de chlorure de cadmium dans l'eutectique LiCl-KCl. Le probleme du choix des electrodes indicatrices a ete particulierement etudie. La methode permet l'analyse quantitative des solutions fondues ainsi que la determination des coefficients de diffusion et de leurs energies d'activation: elle donne en outre des renseignements sur la nature et le comportement des especes ioniques en solution. Elle a ete appliquee a l'etude des solutions des chlorures d'uranium et de plutonium dans l'eutectique LiCl-KCl. La chronoamperometrie lineaire a ete utilisee pour l'etude qualitative de ces solutions, mais la chronopotentiometrie, dont la theorie est plus complete, convient mieux pour l'etude quantitative. Les resultats obtenus ont permis de determiner les

  18. EURADOS intercomparison on measurements and Monte Carlo modelling for the assessment of Americium in a USTUR leg phantom

    International Nuclear Information System (INIS)

    A collaboration of the EURADOS working group on 'Internal Dosimetry' and the United States Transuranium and Uranium Registries (USTUR) has taken place to carry out an intercomparison on measurements and Monte Carlo modelling determining americium deposited in the bone of a USTUR leg phantom. Preliminary results and conclusions of this intercomparison exercise are presented here. (authors)

  19. Plutonium Proliferation: The Achilles Heel of Disarmament

    Energy Technology Data Exchange (ETDEWEB)

    Leventhal, Paul (President, Nuclear Control Institute, Washington D.C.)

    2001-02-07

    Plutonium is a byproduct of nuclear fission, and it is produced at the rate of about 70 metric tons a year in the world's nuclear power reactors. Concerns about civilian plutonium ran high in the 1970s and prompted enactment of the Nuclear Non-Proliferation Act of 1978 to give the United States a veto over separating plutonium from U.S.-supplied uranium fuel. Over the years, however, so-called reactor-grade plutonium has become the orphan issue of nuclear non-proliferation, largely as a consequence of pressures from plutonium-separating countries. The demise of the fast breeder reactor and the reluctance of utilities to introduce plutonium fuel in light-water reactors have resulted in large surpluses of civilian, weapons-usable plutonium, which now approach in size the 250 tons of military plutonium in the world. Yet reprocessing of spent fuel for recovery and use of plutonium proceeds apace outside the United States and threatens to overwhelm safeguards and security measures for keeping this material out of the hands of nations and terrorists for weapons. A number of historical and current developments are reviewed to demonstrate that plutonium commerce is undercutting efforts both to stop the spread of nuclear weapons and to work toward eliminating existing nuclear arsenals. These developments include the breakdown of U.S. anti-plutonium policy, the production of nuclear weapons by India with Atoms-for-Peace plutonium, the U.S.-Russian plan to introduce excess military plutonium as fuel in civilian power reactors, the failure to include civilian plutonium and bomb-grade uranium in the proposed Fissile Material Cutoff Treaty, and the perception of emerging proliferation threats as the rationale for development of a ballistic missile defense system. Finally, immobilization of separated plutonium in high-level waste is explored as a proliferation-resistant and disarmament-friendly solution for eliminating excess stocks of civilian and military plutonium.

  20. Plutonium fuel program

    International Nuclear Information System (INIS)

    The project is concerned with developing an advanced method to produce nuclear reactor fuels. Since 1968 EIR has worked successfully on the production of uranium-plutonium mixed carbide using wet gelation chemistry. An important part of the development is irradiating the fuel in materials test reactors and evaluating its performance. During 1979 the programme continued with principal activities of fuel fabrication development, preparation for irradiation testing, performance evaluation, and modelling and plant engineering. (Auth.)

  1. Plutonium controversy

    Energy Technology Data Exchange (ETDEWEB)

    Richmond, C.R.

    1980-01-01

    The toxicity of plutonium is discussed, particularly in relation to controversies surrounding the setting of radiation protection standards. The sources, amounts of, and exposure pathways of plutonium are given and the public risk estimated. (ACR)

  2. In-line monitor of plutonium holdup in glovebox filters

    International Nuclear Information System (INIS)

    An in-line filter holdup monitoring system has been designed and installed in the Los Alamos Scientific Laboratory (LASL) Plutonium Processing Facility to detect plutonium buildup in a glovebox exhaust filter. The filter is located on top of a glovebox in which plutonium and uranium oxide and carbon are blended, milled, and prepared for making advanced fast breeder reactor (FBR) fuel

  3. Weapons-grade plutonium dispositioning. Volume 4. Plutonium dispositioning in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, J.W.; Olsen, C.S.; Sinha, U.P.

    1993-06-01

    This study is in response to a request by the Reactor Panel Subcommittee of the National Academy of Sciences (NAS) Committee on International Security and Arms Control (CISAC) to evaluate the feasibility of using plutonium fuels (without uranium) for disposal in existing conventional or advanced light water reactor (LWR) designs and in low temperature/pressure LWR designs that might be developed for plutonium disposal. Three plutonium-based fuel forms (oxides, aluminum metallics, and carbides) are evaluated for neutronic performance, fabrication technology, and material and compatibility issues. For the carbides, only the fabrication technologies are addressed. Viable plutonium oxide fuels for conventional or advanced LWRs include plutonium-zirconium-calcium oxide (PuO{sub 2}-ZrO{sub 2}-CaO) with the addition of thorium oxide (ThO{sub 2}) or a burnable poison such as erbium oxide (Er{sub 2}O{sub 3}) or europium oxide (Eu{sub 2}O{sub 3}) to achieve acceptable neutronic performance. Thorium will breed fissile uranium that may be unacceptable from a proliferation standpoint. Fabrication of uranium and mixed uranium-plutonium oxide fuels is well established; however, fabrication of plutonium-based oxide fuels will require further development. Viable aluminum-plutonium metallic fuels for a low temperature/pressure LWR include plutonium aluminide in an aluminum matrix (PuAl{sub 4}-Al) with the addition of a burnable poison such as erbium (Er) or europium (Eu). Fabrication of low-enriched plutonium in aluminum-plutonium metallic fuel rods was initially established 30 years ago and will require development to recapture and adapt the technology to meet current environmental and safety regulations. Fabrication of high-enriched uranium plate fuel by the picture-frame process is a well established process, but the use of plutonium would require the process to be upgraded in the United States to conform with current regulations and minimize the waste streams.

  4. Electrochemical behaviour of uranium ions at liquid Bi electrode in molten LiCI-KCl eutectic

    International Nuclear Information System (INIS)

    Pyrochemical reprocessing method based on molten salt electrorefining is ideally suited for separating the fuel elements from the irradiated metal fuel leaving behind the fission products. In the electrorefiner, pure uranium is collected on the solid mandrel cathode and a mixture of uranium, plutonium, americium, neptunium, and curium are collected at a liquid cadmium cathode. Some amount of rare-earth fission products also deposit along with actinides on cadmium cathode due to their similar chemical behaviour as actinides, thereby making the product less pure. Some work has been in progress to explore the use of other reactive cathodes such as bismuth, zinc, aluminium, etc., so as to achieve better separation of actinides from lanthanide fission products. In the present work, U(lII) ion reduction on liquid bismuth electrode in LiCI-KCI-UCI3 electrolyte salt and the thermodynamic properties of U-Bi intermetallics have been discussed. The studies were carried out using cyclic voltammetry and open circuit potentiometry. Under potential deposition of uranium takes place on liquid bismuth cathode as the activity of uranium is reduced in bismuth cathode due to formation of stable intermetallic compounds. The equilibrium potential for the redox process was evaluated in the temperature range 660-810 K. (author)

  5. An evaluation of the dissolution process of natural uranium ore as an analogue of nuclear fuel

    International Nuclear Information System (INIS)

    The assumption of congruent dissolution of uraninite as a mechanism for the dissolution behaviour of spent fuel was critically examined with regard to the fate of toxic radionuclides. The fission and daughter products of uranium are typically present in spent unreprocessed fuel rods in trace abundances. The principles of trace element geochemistry were applied in assessing the behaviour of these radionuclides during fluid/solid interactions. It is shown that the behaviour of radionuclides in trace abundances that reside in the crystal structure can be better predicted from the ionic properties of these nuclides rather than from assuming that they are controlled by the dissolution of uraninite. Geochemical evidence from natural uranium ore deposits (Athabasca Basin, Northern Territories of Australia, Oklo) suggests that in most cases the toxic radionuclides are released from uraninite in amounts that are independent of the solution behaviour of uranium oxide. Only those elements that have ionic and thus chemical properties similar to U4+, such as plutonium, americium, cadmium, neptunium and thorium can be satisfactorily modelled by the solution properties of uranium dioxide and then only if the environment is reducing. (84 refs., 7 tabs.)

  6. Monitored Natural Attenuation of Inorganic Contaminants in Ground Water Volume 3 Assessment for Radionuclides IncludingTritium, Radon, Strontium, Technetium, Uranium, Iodine, Radium, Thorium, Cesium, and Plutonium-Americium

    Science.gov (United States)

    The current document represents the third volume of a set of three volumes that address the technical basis and requirements for assessing the potential applicability of MNA as part of a ground-water remedy for plumes with nonradionuclide and/or radionuclide inorganic contamina...

  7. Pretreatment of americium/curium solutions for vitrification

    International Nuclear Information System (INIS)

    Vitrification will be used to stabilize an americium/curium (Am/Cm) solution presently stored in F-Canyon for eventual transport to the heavy isotope programs at Oak Ridge National Laboratory. Prior to vitrification, an in-tank oxalate precipitation and a series of oxalic/nitric acid washes will be used to separate these elements and lanthanide fission products from the bulk of the uranium and metal impurities present in the solution. Pretreatment development experiments were performed to understand the behavior of the lanthanides and the metal impurities during the oxalate precipitation and properties of the precipitate slurry. The results of these experiments will be used to refine the target glass composition allowing optimization of the primary processing parameters and design of the solution transfer equipment

  8. Simulation study for the purification of depleted uranium product in FBTR fuel reprocessing

    International Nuclear Information System (INIS)

    A method is developed for the purification of depleted uranium product obtained after partitioning of uranium and plutonium in the third cycle of FBTR fuel reprocessing. Uranium and plutonium were partitioned and recovered by AUC method. It is observed from the study that plutonium recovery is quantitative (100%) if Pu concentration is 0.6 g/L. (author)

  9. Characterization of the deviation of the ideality of concentrated electrolytic solutions: plutonium 4 and uranium 4 nitrate salts study; Contribution a la caracterisation de l'ecart a l'idealite des solutions concentrees d'electrolytes: application aux cas de nitrates de plutonium (4) et d'uranium (4)

    Energy Technology Data Exchange (ETDEWEB)

    Charrin, N

    2000-07-01

    The purpose of this work was to establish a new binary data base by compiling the activity coefficients of plutonium and uranium at oxidation state +IV to better account for media effects in the liquid-liquid extraction operations implemented to reprocess spent nuclear fuel. Chapter 1: first reviews the basic thermodynamic concepts before describing the issues involved in acquiring binary data for the tetravalent actinides. The difficulties arise from two characteristics of this type of electrolyte: its radioactive properties (high specific activity requiring nuclearization of the experimental instrumentation) and its physicochemical properties (strong hydrolysis). After defining the notion of fictive binary data, an approach based on the thermodynamic concept of simple solutions is described in which the activity coefficient of an aqueous phase constituent is dependent on two parameters: the water activity of the system and the total concentration of dissolved constituents. The method of acquiring fictive binary electrolyte data is based on water activity measurements for ternary or quaternary actinide mixtures in nitric acid media, and binary data for nitric acid. The experimental value is then correlated with the characteristics of the fictive binary solution of the relevant electrolyte. Chapter 2: proposes more reliable binary data for nitric acid than the published equivalents, the disparities of which are discussed. The validation of the method described in Chapter 1 for acquiring fictive binary data is then addressed. The test electrolyte, for which binary data are available in the literature, is thorium(IV) nitrate. The method is validated by comparing the published binary data obtained experimentally for binary solutions with the data determined for the ternary Th(NO{sub 3}){sub 4}/HNO{sub 3}/H{sub 2}O system investigated in this study. The very encouraging results of this comparison led us to undertake further research in this area. Chapter 3 discusses

  10. 石墨晶体预衍射X射线荧光法同时测定工艺样品中的低浓铀和低浓钚%Simultaneous Determination of Low Concentration Uranium and Plutonium in Reprocessing Samples by HOPG Pre-Diffraction EDXRF

    Institute of Scientific and Technical Information of China (English)

    吉永超; 牟凌; 马精德; 姜国杜

    2012-01-01

    An analytical method for low concentration uranium and plutonium was developed. In this method uranium and plutonium sampled from nuclear fuel reprocessing solutions were simultaneously determined by highly oriented pyrolytic graphite (HOPG) pre-diffraction EDXRF. The matrix influence was corrected by plotting I(U)/Ic(Ag)-p(U) and I(Pu)/rc(Ag)-p(Pu) instead of I(U)-m(U) and I(Pu)-m(Pu) as their calibration curves where I(Ag) was the intensity of a scattered line from a silver target. The square of correlation coefficient (r2) is better than 0. 999 0 for both uranium and plutonium in the concentration range 2. 00-1 000. 00 mg/L. The experimental measurement results show that the precision (sr) is better than 5% for both uranium and plutonium, and the relative error is less than 3. 0% and 4. 0% for uranium and plutonium, respectively.%利用石墨晶体预衍射X射线荧光法同时测定后处理工艺料液中的低浓铀和钚,建立了测定工艺料液中低浓铀和钚的分析方法.采用Ag靶散射线校正基体效应,以I(U)/Ic(Ag)-ρ(U)和Ⅰ(Pu)/Ic(Ag)-ρ(Pu)作校正曲线,铀和钚的质量浓度均在2.00~1000.00 mg/L时线性相关系数(r2)均大于0.999 0.测定结果表明U和Pu的精密度(Sr)均优于5%,U的相对误差小于3.0%,Pu的相对误差小于4.0%.

  11. Spectrochemical analysis of curium and americium samples

    International Nuclear Information System (INIS)

    Spectrochemical procedures have been developed to determine impurities in americium and curium samples. The simultaneous separation of many impurity elements from the base material (americium and curium) is carried out with extraction and extraction-chromatographic methods using di-2-ethylhexylphosphoric acid. It is shown that part of the elements are separated with extraction or sorption of americium and curium; the other part with the Talspeak process. Two fractions in the extraction chromatography and three fractions in the extraction separation of americium and curium, containing impurities, are analyzed separately by a.c. or d.c. arc spectrography. To increase the sensitivity of the spectrographic analysis and accelerate the burn-up of impurities from the crater of the carbon electrode bismuth fluoride and sodium chloride were used as chemically active substances. The extraction of impurities from weighed quantities of americium and curium samples of 5 to 10 mg permits the lower limit of determined impurity concentrations to be extended to 1 x 10-4 to 5 x 10-3% m/m. (author)

  12. Towards a Plutonium Market

    International Nuclear Information System (INIS)

    Taking as their starting point the various power reactor construction programmes, the authors estimate in the first part of the paper the production of plutonium in the non-socialist world for the period up to 1980. They also estimate the plutonium requirements of the various fast reactor programmes. For the period up to 1970-71, it is found that United States requirements will be satisfied only by drawing on the United States Atomic Energy Commission's stocks and through the exchange agreement with the United Kingdom. For the rest of the world, resources and requirements will be more or less in equilibrium. From 1971 onwards production will greatly exceed the requirements of fast reactors, which will still only be at the prototype stage. If it is assumed that the large-scale construction of fast reactors will not begin until ten years later, the only outlet for plutonium, for which there will no longer be a guarantee of repurchase by the USAEC in the case of enriched uranium reactors, will consist of recycling in thermal reactors. Storage with a view to re-use in breeders at a more remote and indeterminate date would impose an intolerable financial burden on power reactor operators. The question thus arises whether a collapse in the price of plutonium will occur after 1970. In the second part of the paper the authors attempt to predict the course of such a price collapse, followed by a rise as breeders enter service on a commercial basis. They discuss the effects that the absence of plutonium stocks and die competition of recycling in thermal reactors might have on the commercial introduction of breeders. Finally, an attempt is made to determine by what mechanisms a harmonious market for plutonium could be assured during the next fifteen years. (author)

  13. Studies on the Precise Measurement of Isotope of Trace Plutonium in Uranium by MC-ICP-MS%多接收电感耦合等离子体质谱法精确测量铀中痕量钚同位素方法研究

    Institute of Scientific and Technical Information of China (English)

    李力力; 李金英; 赵永刚; 常志远; 张继龙; 王同兴

    2009-01-01

    The infinitesimal plutonium impurity was contained in the uranium product obtained by reprocessing. The precise measurement of the isotope ratio of plutonium was beneficial to the identification of the corresponding source of uranium product. Basing on the separation of trace plutonium from uranium matrix, the operating parameters of MC-ICP-MS were studied and optimized for determining the isotope ratio of trace plutonium, such as the adjustment of resolution, selection of mode of ion extraction, elimination of background and interference, etc. The correction method of the mass discrimination was explored for MCICP-MS measuring isotope ratio. The isotope ratio of ~(239)pu/~(240)Pu in two selected uranium products obtained by reprocessing, which was measured and corrected by measuring plutonium reference material. The ratio of uranium and plutonium is more than 10~(10), and the sum of relative uncertainty for plutonium measurement is better than 5%. The measuring method of isotope ratiois establishes for ng · L~(-1) level of plutonium.%经过核燃料后处理得到的铀产品中含有极微量钚杂质,对其同位素组成进行分析有利于对应铀材料的溯源.本工作在铀钚化学分离的基础上,对影响MC-ICP-MS法精确测量痕量钚同位素丰度比值的仪器条件进行了优化,研究了仪器分辨率、离子提取模式的选择、"背景和干扰"等参数的影响,探索研究了MC-ICP-MS上质量偏倚的校正方法.对选择的两个后处理铀产品中痕量钚同位素比进行了测量,当铀钚比例达10~(10)时,测得的~(239)Pu/~(240)Pu相对不确定度优于5%.并且建立了在MC-ICP-MS上测量ng·L~(-1)量级钚同位素丰度比值的方法.

  14. The plutonium cycle under high surveillance

    International Nuclear Information System (INIS)

    The plutonium is a source of strong public anxiety because of the risk of uncontrolled nuclear weapons proliferation. Several countries have decided to adopt a politics of transparency by drawing up the inventory of plutonium stocks. Most of the plutonium available comes from the operation of nuclear power plants and from the dismantlement of US and Soviet nuclear weapons. A part of plutonium reserves can be burn in fast breeder reactors while the rest can be recycled to produce the Mox (mixed oxides) fuel made of 5 to 7 % of plutonium and 93 to 95 % of uranium 238. The Mox fuel is presently tested in 20 European reactor from which seven are French. The use of Mox requires some modifications of the reactor conception such as the addition of supplementary control rods. The expectations from now to the year 2000 are to use Mox in 28 French reactors, which represents half of the EdF (Electricite de France) units. The paper describes the Melox facility sited in Marcoule (Gard, France) and devoted to the fabrication of Mox fuel from uranium and plutonium oxides. The paper also focusses on the safety and security aspects of plutonium transportation and storage. In a second part, the plutonium isotopes properties are briefly described together with the possible use of plutonium for the fabrication of radiological weapons in the case of nuclear materials diversion. (J.S.). 1 fig., 5 photos

  15. Design of In-situ Measurement System of Uranium Concentration in LiCl-KCl Eutectic Electrolyte

    International Nuclear Information System (INIS)

    Pyrochemical process technology of nuclear fuels using a LiCl-KCl eutectic, currently being developed in the Korea Atomic Energy Research Institute (KAERI) as promising options for future spent nuclear fuel management in Korea, has many advantages such as compactness, economy, radiation stability of the solvent, and non-proliferation. In this pyrochemical process, it is essential to have in-situ information on the concentration of process material such as uranium, plutonium, and fission products as well as their chemical and physical behaviors in a high temperature molten salt medium, since the actinides and lanthanides in the molten salt play an important role for an effective actinide metal recovery during the electro-refining step. In the electro-refining step of the pyrochemical process, a small amount of UCl3 is added to a LiCl-KCl eutectic bath to initiate electro-deposition of uranium onto a working electrode. Since the concentration of uranium in the melts has a tendency to decrease after several batches of process, the concentration of uranium has to be checked continuously, especially when being transferred to the electro-winning step to collect the minor actinides such as neptunium, plutonium, americium, and curium left over in the salts. The study of the application of a spectroscopic technique, especially absorption spectroscopy, which is known as the one of the most powerful tools to apply on-line monitoring, has been continuously increased for the investigation of the behavior of RE elements as well as actinides present in molten salt. In this report, the applicability of UV-Vis absorption spectroscopy was studied as the on-line monitoring method for uranium concentrations in molten chloride electrolyte and a new design of UV-Vis absorption device setup was proposed for direct application to the electro-refining process

  16. Radiochemical determination of Np-237 in soil samples contaminated with weapon grade plutonium

    International Nuclear Information System (INIS)

    The Palomares terrestrial ecosystem (Andalusia, southwestern Spain) is known to constitute a natural laboratory to study the distribution, behaviour and migration of certain actinides, such as plutonium, americium and neptunium. This scenario is partially contaminated with weapon grade plutonium since the burn-out and fragmentation of two of the four thermonuclear bombs accidentally dropped by a B-52 from the USA Air Force back in 1966. While performing radiometric measurements on the field, with the goal of gathering information about the surface contamination levels, the possible presence of 237Np was observed through its 29 keV gamma emission. To accomplish a more detailed characterization of the source term in the contaminated area using the isotopic ratios Pu-Am-Np, the radiochemical isolation and later quantification by alpha spectrometry of 237Np was initiated. As a first approach, a bibliographic study was undertaken, considering different radiochemical methods applied to a variety of samples such as soils, sediments, sea water, lichens, etc. The radiochemical procedure selected in our laboratory involves separation of neptunium from americium, uranium and plutonium with ionic resin (AG 1x2), given that in soil samples from Palomares 239+240Pu levels are several orders of magnitude higher than 237Np. Then neptunium is isolated using TEVA organic resin. After electro-deposition, quantification is performed by high resolution alpha spectrometry. Different analyses have been performed with blank solutions spiked with 236Pu, 237Np and 244Cm, solutions resulting from the total dissolution of isolated radioactive particles and soil samples. The lack of an appropriate tracer in our lab led to the determination of an average percentage of Np recovery using a certified solution of 237Np. Decontamination percentages obtained during the Pu-Np separation ranged from 98 % to 100 %. Some tests to investigate the effect of the addition or absence of NaNO2 (responsible for

  17. Plutonium Finishing Plant safety evaluation report

    International Nuclear Information System (INIS)

    The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE's independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91

  18. Plutonium Finishing Plant safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

    The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE`s independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91.

  19. Determination of plutonium isotopes in urine samples from radiation workers using 236Pu tracer, anion exchange resin and alpha spectrometry

    International Nuclear Information System (INIS)

    Bioassay technique is used for the estimation of actinides present in the body based on their excretion rate through body fluids. For occupational radiation workers urine assay is the preferred method for monitoring of chronic internal exposure. Determination of low concentrations of actinides such as plutonium, americium and uranium at low level of mBq in urine by alpha spectrometry requires pre-concentration of large volumes of urine. This paper deals with standardization of analytical method for the determination of Pu-isotopes in urine samples using anion exchange resin and 236Pu tracer for radiochemical recovery. The method involves oxidation of urine followed by co-precipitation of plutonium along with calcium phosphate. Separation of Pu was carried out by Amberlite, IRA-400, anion exchange resin. Pu-fraction was electrodeposited and activity estimated using tracer recovery by alpha spectrometer. Twenty routine urine samples of radiation workers were analyzed and consistent radiochemical tracer recovery was obtained in the range 74-96% with a mean and standard deviation of 85 and 6% respectively. (author)

  20. Interaction and transport of actinides in natural clay rock with consideration of humic substances and clay organics. Characterization and quantification of the influence of clay organics on the interaction and diffusion of uranium and americium in the clay. Joint project

    Energy Technology Data Exchange (ETDEWEB)

    Bernhard, Gert [Helmholtz-Zentrum Dresden-Rossendorf e.V. (Germany). Inst. of Radiochemistry; Schmeide, Katja; Joseph, Claudia; Sachs, Susanne; Steudtner, Robin; Raditzky, Bianca; Guenther, Alix

    2011-07-01

    , was found to be endothermic and entropy-driven. In contrast, the complex stability constants determined for U(VI) humate complexation at 20 and 40 C are comparable, however, decrease at 60 C. For aqueous U(IV) citrate, succinate, mandelate and glycolate species stability constants were determined. These ligands, especially citrate, increase solubility and mobility of U(IV) in solution due to complexation. The U(VI) sorption onto crushed Opalinus Clay (OPA, Mont Terri, Switzerland) was studied in the absence and presence of HA or low molecular weight organic acids, in dependence on temperature and CO2 presence using OPA pore water as background electrolyte. Distribution coefficients (K{sub d}) were determined for the sorption of U(VI) and HA onto OPA with (0.0222 {+-} 0.0004) m{sup 3}/kg and (0.129 {+-} 0.006) m{sup 3}/kg, respectively. The U(VI) sorption is not influenced by HA ({<=}50 mg/L), however, decreased by low molecular weight organic acids ({>=} 1 x 10{sup -5} M), especially by citrate and tartrate. With increasing temperature, the U(VI) sorption increases both in the absence and in the presence of clay organics. The U(VI) diffusion in compacted OPA is not influenced by HA at 25 and 60 C. Predictions of the U(VI) diffusion show that an increase of the temperature to 60 C does not accelerate the migration of U(VI). With regard to uranium-containing waste, it is concluded that OPA is suitable as host rock for a future nuclear waste repository since OPA has a good retardation potential for U(VI). (orig.)

  1. United States transuranium and uranium registries - 25 years of growth, research, and service. Annual report, April 1992--September 1993

    International Nuclear Information System (INIS)

    The Registries originated in 1968 as the National Plutonium Registry with the name changed to the United States Transuranium Registry the following year to reflect a broader concern with the heavier actinides as well. Initially, the scientific effort of the USTR was directed towards study of the distribution and dose of plutonium and americium in occupationally exposed persons, and to assessment of the effects of exposure to the transuranium elements on health. This latter role was reassessed during the 1970's when it was recognized that the biased cohort of the USTR was inappropriate for epidemiologic analysis. In 1978, the administratively separate but parallel United States Uranium Registry was created to carry out similar work among persons exposed to uranium and its decay products. A seven member scientific advisory committee provided guidance and scientific oversight. In 1992, the two Registries were administratively combined and transferred from the purview of a Department of Energy contractor to Washington State University under the provisions of a grant. Scientific results for the first twenty-five years of the Registries are summarized, including the 1985 publication of the analysis of the first whole body donor. Current scientific work in progress is summarized along with administrative activities for the period

  2. United States transuranium and uranium registries - 25 years of growth, research, and service. Annual report, April 1992--September 1993

    Energy Technology Data Exchange (ETDEWEB)

    Kathren, R.L.; Harwick, L.A.; Toohey, R.E.; Russell, J.J.; Filipy, R.E.; Dietert, S.E.; Hunacek, M.M.; Hall, C.A.

    1994-10-01

    The Registries originated in 1968 as the National Plutonium Registry with the name changed to the United States Transuranium Registry the following year to reflect a broader concern with the heavier actinides as well. Initially, the scientific effort of the USTR was directed towards study of the distribution and dose of plutonium and americium in occupationally exposed persons, and to assessment of the effects of exposure to the transuranium elements on health. This latter role was reassessed during the 1970`s when it was recognized that the biased cohort of the USTR was inappropriate for epidemiologic analysis. In 1978, the administratively separate but parallel United States Uranium Registry was created to carry out similar work among persons exposed to uranium and its decay products. A seven member scientific advisory committee provided guidance and scientific oversight. In 1992, the two Registries were administratively combined and transferred from the purview of a Department of Energy contractor to Washington State University under the provisions of a grant. Scientific results for the first twenty-five years of the Registries are summarized, including the 1985 publication of the analysis of the first whole body donor. Current scientific work in progress is summarized along with administrative activities for the period.

  3. Pyrochemical processing of plutonium. Technology review report

    International Nuclear Information System (INIS)

    Non-aqueous processes are now in routine use for direct conversion of plutonium oxide to metal, molten salt extraction of americium, and purification of impure metals by electrorefining. These processes are carried out at elevated temperatures in either refractory metal crucibles or magnesium-oxide ceramics in batch-mode operation. Direct oxide reduction is performed in units up to 700 gram PuO2 batch size with molten calcium metal as the reductant and calcium chloride as the reaction flux. Americium metal is removed from plutonium metal by salt extraction with molten magnesium chloride. Electrorefining is used to isolate impurities from molten plutonium by molten salt ion transport in a controlled potential oxidation-reduction cell. Such cells can purify five or more kilograms of impure metal per 5-day electrorefining cycle. The product metal obtained is typically > 99.9% pure, starting from impure feeds. Metal scrap and crucible skulls are recovered by hydriding of the metallic residues and recovered either as impure metal or oxide feeds

  4. Plutonium credit

    International Nuclear Information System (INIS)

    The report describes problems of long-term commercial contracts for an assured uranium supply, uranium enrichment and reprocessing, or rather the back end of the fuel cycle. Measures to improve these conditions are proposed

  5. United States Transuranium and Uranium Registries. Annual Report, October 1, 1993--September 30, 1994

    International Nuclear Information System (INIS)

    This report summarizes the salient activities and progress of the United States Transuranium. and Uranium Registries for the period October 1, 1993 through September 30, 1994, along with details of specific programs areas including the National Human Radiobiology Tissue Repository (NHRTR) and tissue radiochemistry analysis project. Responsibility for tissue radioanalysis was transferred from Los Alamos National Laboratory to Washington State University in February 1994. The University of Washington was selected as the Quality Assurance/Quality Control laboratory and a three way intercomparison with them and LANL has been initiated. The results of the initial alpha spectrometry intercomparison showed excellent agreement among the laboratories and are documented in full in the Appendices to the report. The NHRTR serves as the initial point of receipt for samples received from participants in the USTUR program. Samples are weighed, divided, and reweighed, and a portion retained by the NHRTR as backup or for use in other studies. Tissue specimens retained in the NHRTR are maintained frozen at -70 C and include not only those from USTUR registrants but also those from the radium dial painter and thorium worker studies formerly conducted by Argonne National Laboratory. In addition, there are fixed tissues and a large collection of histopathology slides from all the studies, plus about 20,000 individual solutions derived from donated tissues. These tissues and tissue related materials are made available to other investigators for legitimate research purposes. Ratios of the concentration of actinides in various tissues have been used to evaluate the biokinetics, and retention half times of plutonium and americium. Retention half times for plutonium in various soft tissues range from 10-20 y except for the testes for which a retention half time of 58 y was observed. For americium, the retention half time in various soft tissues studied was 2.2-3.5 y

  6. United States Transuranium and Uranium Registries. Annual Report, October 1, 1993--September 30, 1994

    Energy Technology Data Exchange (ETDEWEB)

    Kathren, R.L.; Harwick, L.A. [comps.] [eds.

    1995-08-01

    This report summarizes the salient activities and progress of the United States Transuranium. and Uranium Registries for the period October 1, 1993 through September 30, 1994, along with details of specific programs areas including the National Human Radiobiology Tissue Repository (NHRTR) and tissue radiochemistry analysis project. Responsibility for tissue radioanalysis was transferred from Los Alamos National Laboratory to Washington State University in February 1994. The University of Washington was selected as the Quality Assurance/Quality Control laboratory and a three way intercomparison with them and LANL has been initiated. The results of the initial alpha spectrometry intercomparison showed excellent agreement among the laboratories and are documented in full in the Appendices to the report. The NHRTR serves as the initial point of receipt for samples received from participants in the USTUR program. Samples are weighed, divided, and reweighed, and a portion retained by the NHRTR as backup or for use in other studies. Tissue specimens retained in the NHRTR are maintained frozen at -70 C and include not only those from USTUR registrants but also those from the radium dial painter and thorium worker studies formerly conducted by Argonne National Laboratory. In addition, there are fixed tissues and a large collection of histopathology slides from all the studies, plus about 20,000 individual solutions derived from donated tissues. These tissues and tissue related materials are made available to other investigators for legitimate research purposes. Ratios of the concentration of actinides in various tissues have been used to evaluate the biokinetics, and retention half times of plutonium and americium. Retention half times for plutonium in various soft tissues range from 10-20 y except for the testes for which a retention half time of 58 y was observed. For americium, the retention half time in various soft tissues studied was 2.2-3.5 y.

  7. 1976 Hanford americium exposure incident: psychological aspects

    International Nuclear Information System (INIS)

    Accidents involving exposure to radiation or radioactive materials may involve an unusual degree of emotional trauma. Methods that may be employed in dealing with such trauma are discussed in relation to a specific accident in which a radiation worker was injured and seriously contaminated with americium-241

  8. Intake and digestibility of range plants grown on plutonium contaminated soils as determined with grazing cattle

    International Nuclear Information System (INIS)

    Area 13 is one of several areas of the Nevada Test Site contaminated with transuranics. Cattle were grazed on the area to study the botanical and chemical composition of the forage, the digestibility of range plants as selected by range cattle, and the intake of plutonium and americium by grazing cattle. The botanical and chemical composition of the diet of cattle grazing on plutonium-contaminated range was determined. The major portion of the diet was browse plants which were high in fiber and ash but low in energy. Daily feed intake of the grazing animals was also determined so that the amount of nuclides ingested daily could be ascertained. Cattle generally consumed over 2 kilograms per 100 kilograms body weight of dry matter daily which resulted in a daily intake of 3600 to 6600 picocuries of plutonium-238, 85,000 to 400,000 picocuries of plutonium-239, and 11,000 to 31,000 picocuries of americium-241. The soil ingested by range cattle constituted the principal source of ingested plutonium and americium. This is not unexpected as plutonium oxide is one of the least soluble substances known and the range studied is one of very limited rainfall. As expected, the forage from an inner compound was contaminated to a greater extent than the range plants from an outer compound

  9. Plutonium, cesium, uranium and thorium series radionuclides in the Hudson River estuary and other environments. Annual technical progress report, December 1, 1983-November 30, 1984

    International Nuclear Information System (INIS)

    Radionuclide activities were measured in sediment cores and suspended particle samples from the Hudson River estuary. Activities of 137Cs, 134Cs, and 60Co, 239240Pu and 238Pu indicate rapid accumulation in marginal cove areas, and very rapid deposition in the harbor adjacent to New York City, resulting in 239240Pu accumulations of more than an order of magnitude greater than the fallout delivery rate. Fallout 239240Pu moving downstream appears to be retained within the system by particle deposition, while more than 50% of the 137Cs derived from both reactor releases and fallout has been exported. Significant movement of dissolved plutonium into the estuary from adjacent coastal waters may be occurring. Depth profiles of radionuclides are not significantly altered by physical mixing processes in areas accumulating particles at greater than 1 cm/yr. Transport of fallout radionuclides appears to have decreased faster than would be calculated from continuous removal from a well-mixed soil reservoir, indicating that sequestering of a substantial portion of the soil fallout burden has occurred in the watershed soils over the past two decades. Measurements of fallout 239240Pu in a saline lake with a high carbonate ion concentration yielded water column activities two orders of magnitude greater than that found for fallout plutonium in other continental waters, indicating extensive mobility in some natural water environments. Experiments using lake water suggest that carbonate ions are likely to be important in regulating plutonium solubility in some environments and that low molecular weight complexes are primarily responsible for enhanced plutonium solubility. 45 references, 17 figures, 14 tables

  10. Plutonium Chemistry in the UREX Separation Processes

    International Nuclear Information System (INIS)

    The objective of the project is to examine the chemical speciation of plutonium in UREX+ (uranium/tributylphosphate) extraction processes for advanced fuel technology. Researchers will analyze the change in speciation using existing thermodynamics and kinetic computer codes to examine the speciation of plutonium in aqueous and organic phases. They will examine the different oxidation states of plutonium to find the relative distribution between the aqueous and organic phases under various conditions such as different concentrations of nitric acid, total nitrates, or actinide ions. They will also utilize techniques such as X-ray absorbance spectroscopy and small-angle neutron scattering for determining plutonium and uranium speciation in all separation stages. The project started in April 2005 and is scheduled for completion in March 2008.

  11. Plutonium solubilities

    International Nuclear Information System (INIS)

    Thermochemical data has been selected for plutonium oxide, hydroxide, carbonate and phosphate equilibria. Equilibrium constants have been evaluated in the temperature range 0 to 300 degrees C at a pressure of 1 bar to T≤100 degrees C and at the steam saturated pressure at higher temperatures. Measured solubilities of plutonium that are reported in the literature for laboratory experiments have been collected. Solubility data on oxides, hydroxides, carbonates and phosphates have been selected. No solubility data were found at temperatures higher than 60 degrees C. The literature solubility data have been compared with plutonium solubilities calculated with the EQ3/6 geochemical modelling programs, using the selected thermodynamic data for plutonium. (authors)

  12. Plutonium-238 processing at Savannah River Plant

    Energy Technology Data Exchange (ETDEWEB)

    Burney, G.A.

    1983-01-01

    Plutonium-238 is produced by irradiating NpO/sub 2/-Al cermet slugs or tubes with neutrons. The neptunium-237 is produced as a by-product when natural or enriched uranium is irradiated with neutrons. The neptunium is separated by solvent extraction and ion exchange and precipitated as neptunium oxalate. Neptunium oxalate is calcined to neptunium oxide and fabricated into targets for irradiation. The irradiation conditions are controlled to produce plutonium with 80 to 90 wt % /sup 238/Pu.

  13. Interphase distribution of plutonium and strontium in chloride melts in the presence of oxide phases

    International Nuclear Information System (INIS)

    Interphase distribution of plutonium and strontium in the eutectic melt of sodium and potassium chlorides at the 1023 and 1173 K temperature in the presence of solids La2O3 and ZrO2 has been studied to develop the methods for actinoid and fragment element separation by extraction of solid phase elements from molten salts. It is established that 0.5 g of La2O3 during an hour practically extracts the whole plutonium from the salt phase. Under similar conditions ZrO2 extracts plutonium slightly worse. It is shown that the nature of the halogen melt doesn't influence on plutonium and strontium distribution. La2O3 may extract plutonium and americium both from fluoride and chloride melts, and ZrO2 may extract only plutonium. Alkaline and alkaline-earth fragment elements (Cs, Sr) are not extracted by these oxides from chloride melts

  14. Plutonium, cesium and uranium series radionuclides in the Hudson River estuary and other environments. Annual technical progress report, December 1, 1979-November 30, 1980

    International Nuclear Information System (INIS)

    Radionuclide activities were measured in a large number of sediment cores and suspended particle samples throughout the salinity range of the Hudson River estuary. Activities of 137Cs, 134Cs and 60Co determined by gamma spectrometry and 239240Pu and 238Pu determined by alpha spectrometry indicate reasonably rapid accumulation rates in the sediments of marginal cove areas, and very rapid deposition in the harbor region adjacent to New York City, resulting in 239240Pu accumulations there more than an order of magnitude greater than the fallout delivery rate. Measurable amounts of reactor-derived 134Cs and 60Co are found in nearly al sediment samples containing appreciable 137Cs between 15 km upstream of Indian Point and the downstream extent of our sampling about 70 km south of the reactor. Fallout 239240Pu reaching the Hudson appears to be almost completely retained within the systems by particle deposition, while 70 to 90% of the 137Cs derived from both reactor releases and fallout has been exported to the coastal waters in solution. Activity levels of 239240Pu in New York harbor sediments indicate a significant source in addition to suspended particles carried down the Hudson. The most likely cause appears to be transport into the estuary of particles from offshore waters having higher specific activities of 239240Pu. Measurements of fallout 239240Pu in a saline lake with a high carbonate ion concentration yielded water column activities about two orders of magnitude greater than has been found for fallout plutonium in other continental waters, indicating extensive mobility in some natural water environments. Experiments using lake water suggest that carbonate ion may indeed be a critical factor in regulating plutonium solubility and that low molecular weight complexes are primarily responsible for enhanced plutonium solubility

  15. Plutonium, cesium and uranium series radionuclides in the Hudson River estuary and other environments. Annual technical progress report, December 1, 1979-November 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, H. J.; Trier, R. M.; Olsen, C. R.

    1980-01-01

    Radionuclide activities were measured in a large number of sediment cores and suspended particle samples throughout the salinity range of the Hudson River estuary. Activities of /sup 137/Cs, /sup 134/Cs and /sup 60/Co determined by gamma spectrometry and /sup 239/ /sup 240/Pu and /sup 238/Pu determined by alpha spectrometry indicate reasonably rapid accumulation rates in the sediments of marginal cove areas, and very rapid deposition in the harbor region adjacent to New York City, resulting in /sup 239/ /sup 240/Pu accumulations there more than an order of magnitude greater than the fallout delivery rate. Measurable amounts of reactor-derived /sup 134/Cs and /sup 60/Co are found in nearly al sediment samples containing appreciable /sup 137/Cs between 15 km upstream of Indian Point and the downstream extent of our sampling about 70 km south of the reactor. Fallout /sup 239/ /sup 240/Pu reaching the Hudson appears to be almost completely retained within the systems by particle deposition, while 70 to 90% of the /sup 137/Cs derived from both reactor releases and fallout has been exported to the coastal waters in solution. Activity levels of /sup 239/ /sup 240/Pu in New York harbor sediments indicate a significant source in addition to suspended particles carried down the Hudson. The most likely cause appears to be transport into the estuary of particles from offshore waters having higher specific activities of /sup 239/ /sup 240/Pu. Measurements of fallout /sup 239/ /sup 240/Pu in a saline lake with a high carbonate ion concentration yielded water column activities about two orders of magnitude greater than has been found for fallout plutonium in other continental waters, indicating extensive mobility in some natural water environments. Experiments using lake water suggest that carbonate ion may indeed be a critical factor in regulating plutonium solubility and that low molecular weight complexes are primarily responsible for enhanced plutonium solubility.

  16. Nuclear legacy. Democracy in a plutonium economy

    International Nuclear Information System (INIS)

    There have already been a few hundred known incidents of nuclear smuggling, mostly of small quantities not close to weapons grade material - but one gram of plutonium is more than sufficient to cause significant harm and to pose a substantial threat. The potential for further thefts is growing as the world produces ever more quantities of plutonium, not only from the dismantling of nuclear weapons but also from the separation out of plutonium from spent uranium nuclear reactor fuel elements. Trying to prevent the theft of gram quantities of plutonium would require levels of protection and surveillance unacceptably high in a democratic society. It is unlikely, therefore, that democracy could survive in a plutonium economy

  17. Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 415: Project 57 No. 1 Plutonium Dispersion (NTTR), Nevada, Revision 0

    Energy Technology Data Exchange (ETDEWEB)

    Matthews, Patrick; Burmeister, Mark

    2014-04-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the actions needed to achieve closure for Corrective Action Unit (CAU) 415, Project 57 No. 1 Plutonium Dispersion (NTTR). CAU 415 is located on Range 4808A of the Nevada Test and Training Range (NTTR) and consists of one corrective action site: NAFR-23-02, Pu Contaminated Soil. The CAU 415 site consists of the atmospheric release of radiological contaminants to surface soil from the Project 57 safety experiment conducted in 1957. The safety experiment released plutonium (Pu), uranium (U), and americium (Am) to the surface soil over an area of approximately 1.9 square miles. This area is currently fenced and posted as a radiological contamination area. Vehicles and debris contaminated by the experiment were subsequently buried in a disposal trench within the surface-contaminated, fenced area and are assumed to have released radiological contamination to subsurface soils. Potential source materials in the form of pole-mounted electrical transformers were also identified at the site and will be removed as part of closure activities.

  18. Uses of depleted uranium

    International Nuclear Information System (INIS)

    The depleted uranium is that in which percentage of uranium-235 fission executable is less than 0.2% or 0.3%. It is usually caused by the process of reprocessing the nuclear fuel burning, and also mixed with some other radioactive elements such as uranium 236, 238 and plutonium 239. The good features of the depleted uranium are its high density, low price and easily mined. So, the specifications for depleted uranium make it one of the best materials in case you need to have objects small in size, but quite heavy regarding its size. Uses of deplet ed uranium were relatively increased in domestic industrial uses as well as some uses in nuclear industry in the last few years. So it has increased uses in many areas of military and peaceful means such as: in balancing the giant air crafts, ships and missiles and in the manufacture of some types of concrete with severe hardness. (author)

  19. Fuel cycles using adulterated plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Brooksbank, R. E.; Bigelow, J. E.; Campbell, D. O.; Kitts, F. G.; Lindauer, R. B.

    1978-01-01

    Adjustments in the U-Pu fuel cycle necessitated by decisions made to improve the nonproliferation objectives of the US are examined. The uranium-based fuel cycle, using bred plutonium to provide the fissile enrichment, is the fuel system with the highest degree of commercial development at the present time. However, because purified plutonium can be used in weapons, this fuel cycle is potentially vulnerable to diversion of that plutonium. It does appear that there are technologically sound ways in which the plutonium might be adulterated by admixture with /sup 238/U and/or radioisotopes, and maintained in that state throughout the fuel cycle, so that the likelihood of a successful diversion is small. Adulteration of the plutonium in this manner would have relatively little effect on the operations of existing or planned reactors. Studies now in progress should show within a year or two whether the less expensive coprocessing scheme would provide adequate protection (coupled perhaps with elaborate conventional safeguards procedures) or if the more expensive spiked fuel cycle is needed as in the proposed civex pocess. If the latter is the case, it will be further necessary to determine the optimum spiking level, which could vary as much as a factor of a billion. A very basic question hangs on these determinations: What is to be the nature of the recycle fuel fabrication facilities. If the hot, fully remote fuel fabrication is required, then a great deal of further development work will be required to make the full cycle fully commercial.

  20. Separation of trace elemental impurities from plutonium oxide matrics by solvent extraction

    International Nuclear Information System (INIS)

    Plutonium oxide and uranium oxide are the main products of the PUREX process. Plutonium oxide and uranium oxide are the source materials used in the manufacture of ceramic grade oxide fuels for nuclear reactors. Among the various chemical specifications for plutonium oxide, trace elemental impurities are one of them. Presently, in the PREFRE laboratory, trace elemental assay in PuO2 by ICP-AES is carried out using this Pu-matrix separation method

  1. Study of physico-chemical release of uranium and plutonium oxides during the combustion of polycarbonate and of ruthenium during the combustion of solvents used in the reprocessing of nuclear fuel

    International Nuclear Information System (INIS)

    The level of consequences concerning a fire in a nuclear facility is in part estimated by the quantities and the physico-chemical forms of radioactive compounds that may be emitted out of the facility. It is therefore necessary to study the contaminant release from the fire. Because of the multiplicity of the scenarios, two research subjects were retained. The first one concerns the study of the uranium or plutonium oxides chemical release during the combustion of the polycarbonate glove box sides. The second one is about the physico chemical characterisation of the ruthenium release during the combustion of an organic solvent mixture (tributyl phosphate-dodecane) used for the nuclear fuel reprocessing. Concerning the two research subjects, the chemical release, i.e. means the generation of contaminant compounds gaseous in the fire, was modelled using thermodynamical simulations. Experiments were done in order to determine the ruthenium release factor during solvent combustion. A cone calorimeter was used for small scale experiments. These results were then validated by large scale tests under conditions close to the industrial process. Thermodynamical simulations, for the two scenarios studied. Furthermore, the experiments on solvent combustion allowed the determination of a suitable ruthenium release factor. Finally, the mechanism responsible of the ruthenium release has been found. (author)

  2. Accumulation of polonium 210Po, uranium (234U and 238U and plutonium (238Pu and 239+240Pu in tissues and organs of deer Carvidae from northern Poland

    Directory of Open Access Journals (Sweden)

    Strumińska-Parulska D. I.

    2013-04-01

    Full Text Available Livers, kidneys and muscles of large herbivorous animals: roe deer (Capreolus capreolus, red deer (Cervus elaphus and fallow deer (Dama dama, collected in northern Poland, were the subject of the investigation. Polonium, uranium and plutonium activities were measured using alpha spectrometry along with relevant radiochemical procedures. The average concentrations of 210Po ranged between 0.02±0.01 Bq kg−1 ww in muscles and 7.15±0.12 Bq kg−1 ww in kidneys. The concentration of 238U ranged widely between 2.49±0.18 and 69.37±5.45 mBq kg−1 ww. The average concentrations of 239+240Pu ranged between 0.18±0.04 mBq kg−1 ww in muscle and 2.77±0.60 mBq kg−1 ww in kidneys. Statistical analysis showed that the sampling location, sex, age and species did not influence on the radioisotope activities in the analyzed samples.

  3. A case of internal contamination with plutonium oxide

    International Nuclear Information System (INIS)

    A description is given of a case of plutonium and americium internal contamination due to an accidental glove-box explosion involving plutonium dioxide powder at the Casaccia Plutonium Plant in 1974. The person received a small contaminated wound to the face, a diffused contamination on the hair and considerable activity in the nose. Details are given of the actions taken to reduce the initial contamination and to obtain the maximum information on the residual contamination and on the dose commitment. Results were obtained for 1) the levels of activity in the wound and hair both before and after DTPA treatment; 2) direct lung counting performed at intervals up to 70 days after the incident; 3) the urinary and faecal excretion of plutonium and americium for up to 140 days after contamination; and 4) the levels of activity in blood. Using the data supplied by the lung counting and the excretion curves, the committed dose equivalents were calculated to be lung 1.2 + 4.8 rem, bone 40 rem, liver 100 rem and kidneys 3 rem. The effective total body committed dose equivalent was 7.5 + 8 rem. (U.K.)

  4. Reduction of Worldwide Plutonium Inventories Using Conventional Reactors and Advanced Fuels: A Systems Study

    Energy Technology Data Exchange (ETDEWEB)

    Krakowski, R.A., Bathke, C.G.

    1997-12-31

    The potential for reducing plutonium inventories in the civilian nuclear fuel cycle through recycle in LWRs of a variety of mixed oxide forms is examined by means of a cost based plutonium flow systems model. This model emphasizes: (1) the minimization of separated plutonium; (2) the long term reduction of spent fuel plutonium; (3) the optimum utilization of uranium resources; and (4) the reduction of (relative) proliferation risks. This parametric systems study utilizes a globally aggregated, long term (approx. 100 years) nuclear energy model that interprets scenario consequences in terms of material inventories, energy costs, and relative proliferation risks associated with the civilian fuel cycle. The impact of introducing nonfertile fuels (NFF,e.g., plutonium oxide in an oxide matrix that contains no uranium) into conventional (LWR) reactors to reduce net plutonium generation, to increase plutonium burnup, and to reduce exo- reactor plutonium inventories also is examined.

  5. Recent uranium supply-demand situation and market research in the world

    International Nuclear Information System (INIS)

    The uranium price (spot price) is 20 $ /1 pond U3O8 in September, 2004, which makes a new record since 20 years. By 'Uranium: Resource, Production, Demand' (RED BOOK) (2003) arranged by OECD/NEA and IAEA, uranium resource in the world is able to supply for 85 years without reprocessing and recycle. Accordingly, nuclear power development will not be controlled by uranium resource. However, uranium producer dose not develop new mine in order to fear stock of Power Company and the large amount of secondary source of supply such as enriched uranium. Now, the production rate of uranium mines in the world is about half of demand. If the situation will continue, the uranium price should become high price. Uranium demand, uranium resource, production and supply, stock, enriched uranium from Russia, excess enriched uranium in USA, plutonium, recovery of uranium, excess plutonium by disarmament, depleted uranium and uranium market are discussed. (S.Y.)

  6. Strategies for denaturing the weapons-grade plutonium stockpile

    Energy Technology Data Exchange (ETDEWEB)

    Buckner, M.R.; Parks, P.B.

    1992-10-01

    In the next few years, approximately 50 metric tons of weapons-grade plutonium and 150 metric tons of highly-enriched uranium (HEU) may be removed from nuclear weapons in the US and declared excess. These materials represent a significant energy resource that could substantially contribute to our national energy requirements. HEU can be used as fuel in naval reactors, or diluted with depleted uranium for use as fuel in commercial reactors. This paper proposes to use the weapons-grade plutonium as fuel in light water reactors. The first such reactor would demonstrate the dual objectives of producing electrical power and denaturing the plutonium to prevent use in nuclear weapons.

  7. Study of physico-chemical release of uranium and plutonium oxides during the combustion of polycarbonate and of ruthenium during the combustion of solvents used in the reprocessing of nuclear fuel; Etude de la mise en suspension physico-chimique des oxydes de plutonium et d'uranium lors de la combustion de polycarbonate et de ruthenium lors de la combustion des solvants de retraitement du combustible irradie

    Energy Technology Data Exchange (ETDEWEB)

    Bouilloux, L

    1998-07-01

    The level of consequences concerning a fire in a nuclear facility is in part estimated by the quantities and the physico-chemical forms of radioactive compounds that may be emitted out of the facility. It is therefore necessary to study the contaminant release from the fire. Because of the multiplicity of the scenarios, two research subjects were retained. The first one concerns the study of the uranium or plutonium oxides chemical release during the combustion of the polycarbonate glove box sides. The second one is about the physico chemical characterisation of the ruthenium release during the combustion of an organic solvent mixture (tributyl phosphate-dodecane) used for the nuclear fuel reprocessing. Concerning the two research subjects, the chemical release, i.e. means the generation of contaminant compounds gaseous in the fire, was modelled using thermodynamical simulations. Experiments were done in order to determine the ruthenium release factor during solvent combustion. A cone calorimeter was used for small scale experiments. These results were then validated by large scale tests under conditions close to the industrial process. Thermodynamical simulations, for the two scenarios studied. Furthermore, the experiments on solvent combustion allowed the determination of a suitable ruthenium release factor. Finally, the mechanism responsible of the ruthenium release has been found. (author)

  8. Surface complexation modeling of americium sorption onto volcanic tuff.

    Science.gov (United States)

    Ding, M; Kelkar, S; Meijer, A

    2014-10-01

    Results of a surface complexation model (SCM) for americium sorption on volcanic rocks (devitrified and zeolitic tuff) are presented. The model was developed using PHREEQC and based on laboratory data for americium sorption on quartz. Available data for sorption of americium on quartz as a function of pH in dilute groundwater can be modeled with two surface reactions involving an americium sulfate and an americium carbonate complex. It was assumed in applying the model to volcanic rocks from Yucca Mountain, that the surface properties of volcanic rocks can be represented by a quartz surface. Using groundwaters compositionally representative of Yucca Mountain, americium sorption distribution coefficient (Kd, L/Kg) values were calculated as function of pH. These Kd values are close to the experimentally determined Kd values for americium sorption on volcanic rocks, decreasing with increasing pH in the pH range from 7 to 9. The surface complexation constants, derived in this study, allow prediction of sorption of americium in a natural complex system, taking into account the inherent uncertainty associated with geochemical conditions that occur along transport pathways.

  9. Feasibility Study of 1/3 Thorium-Plutonium Mixed Oxide Core

    OpenAIRE

    2014-01-01

    Thorium-plutonium mixed oxide (Th-MOX) fuel has become one of the most promising solutions to reduce a large and increasing plutonium stockpile. Compared with traditional uranium-plutonium mixed oxide (U-MOX) fuels, Th-MOX fuel has higher consumption rate of plutonium in LWRs. Besides, thorium based fuels have improved thermomechanical material properties compared with traditional U-MOX fuels. Previous studies on a full Th-MOX core have shown reduced efficiency in reactivity control mechanism...

  10. Elicitation of dissolution rate data for potential wasteform types for plutonium under repository conditions

    OpenAIRE

    Deissmann, Guido; Neumeier, Stefan; Brandt, Felix; Modolo, Giuseppe; Bosbach, Dirk

    2013-01-01

    Various plutonium isotopes are generated during the operation of nuclear reactors from the uranium present in the nuclear fuels through capture of neutrons. The plutonium contained within spent nuclear fuels can be recovered during reprocessing. At present, separated stocks of UK civil plutonium (about 90 tHM as PuO2) are held in storage as zero value asset, since there is no final decision on plutonium disposition in the UK. Recently, the UK government stated that its preliminary preferred p...

  11. Individual economical value of plutonium isotopes and analysis of the reprocessing of irradiated fuel

    International Nuclear Information System (INIS)

    An economical analysis of plutonium recycle in a PWR reactor, without any modification, is done, supposing an open market for the plutonium. The individual value of the plutonium isotopes is determined solving a system with four equations, which the unknow factors are the Pu-239, Pu-240, pu-241 and Pu-242 values. The equations are obtained equalizing the cost of plutonium fuel cycle of four different isotope mixture to the cost of the uranium fuel cycle. (E.G.)

  12. Disposal of Surplus Weapons Grade Plutonium

    International Nuclear Information System (INIS)

    The Office of Fissile Materials Disposition is responsible for disposing of inventories of surplus US weapons-usable plutonium and highly enriched uranium as well as providing, technical support for, and ultimate implementation of, efforts to obtain reciprocal disposition of surplus Russian plutonium. On January 4, 2000, the Department of Energy issued a Record of Decision to dispose of up to 50 metric tons of surplus weapons-grade plutonium using two methods. Up to 17 metric tons of surplus plutonium will be immobilized in a ceramic form, placed in cans and embedded in large canisters containing high-level vitrified waste for ultimate disposal in a geologic repository. Approximately 33 metric tons of surplus plutonium will be used to fabricate MOX fuel (mixed oxide fuel, having less than 5% plutonium-239 as the primary fissile material in a uranium-235 carrier matrix). The MOX fuel will be used to produce electricity in existing domestic commercial nuclear reactors. This paper reports the major waste-package-related, long-term disposal impacts of the two waste forms that would be used to accomplish this mission. Particular emphasis is placed on the possibility of criticality. These results are taken from a summary report published earlier this year

  13. Plutonium use - Present status and prospects

    International Nuclear Information System (INIS)

    The use of plutonium in thermal and fast reactors is a demonstrated, if not proven, technology. Moreover, plutonium is being produced in increasing quantities. Evaluation of this production on a world scale shows that it would be theoretically possible to construct numerous breeders and thus to make the best use of plutonium, while considerably reducing uranium consumption. This source of plutonium is nevertheless dependent on the reprocessing of irradiated fuel. Long delays in installing and adequate world reprocessing capacity are weakening the prospects for introducing breeders. Furthermore, the critical situation regarding reprocessing may delay the development of complementary reprocessing methods for fuels with a high plutonium content and high burnup. The recycling of plutonium is now a well-known technique and any objections to it hardly bear analysis. Utilization of plutonium offers an appreciable saving in terms of uranium and separative work units; and it can also be shown that immediate reprocessing of the recycling fuel is not essential for the economics of the concept. Temporary storage of recycled fuel is a particularly safe form of concentrating plutonium, namely in irradiated plutonium-bearing fuel assemblies. Finally, recycling offers such flexibility that it represents no obstacle to fuel management at power plants with light-water reactors. These strategic considerations imply that the technology of using plutonium for fabricating thermal or fast reactor fuels is both technically reliable and economically viable. The methods used in industrial facilities are fully reassuring in this respect. Although various unsolved problems exist, none seems likely to impede current developments, while the industrial experience gained has enabled the economics and reliability of the methods to be improved appreciably. Apart from the techno-economic aspects, the plutonium industry must face extremely important problems in connection with the safety of personnel

  14. Plutonium's human guinea pigs

    International Nuclear Information System (INIS)

    In 1993, the inquiry of an American journalist, Eileen Welson, has revealed to the general public that in the 1940s and 1950s people were used as human samples to study the effects of radiation, without their knowledge. The shock, occurred in American opinion has led to the creation of an inquiry commission, which keeps revealing new documents. The latter shows the existence of so many hundreds of experiments of this type carried out in a spirit too close to that of the Nazi physicians. Injections of plutonium sub 2 sub 3 sub 8 and sub 2 sub 3 sub 9 or enriched uranium, exposition to radioactive clouds, massive irradiation by x-rays,... etc. those practice have persisted till the 1970s. (author). 22 refs., 10 figs

  15. Structural Properties and Charge Distribution of the Sodium Uranium, Neptunium, and Plutonium Ternary Oxides: A Combined X-ray Diffraction and XANES Study.

    Science.gov (United States)

    Smith, Anna L; Martin, Philippe; Prieur, Damien; Scheinost, Andreas C; Raison, Philippe E; Cheetham, Anthony K; Konings, Rudy J M

    2016-02-15

    The charge distributions in α-Na2UO4, Na3NpO4, α-Na2NpO4, Na4NpO5, Na5NpO6, Na2PuO3, Na4PuO5, and Na5PuO6 are investigated in this work using X-ray absorption near-edge structure (XANES) spectroscopy at the U-L3, Np-L3, and Pu-L3 edges. In addition, a Rietveld refinement of monoclinic Na2PuO3, in space group C2/c, is reported for the first time, and the existence of the isostructural Na2NpO3 phase is revealed. In contrast to measurements in solution, the number of published XANES data for neptunium and plutonium solid phases with a valence state higher than IV is very limited. The present results cover a wide range of oxidation states, namely, IV to VII, and can serve as reference for future investigations. The sodium actinide series show a variety of local coordination geometries, and correlations between the shape of the XANES spectra and the local structural environments are discussed herein. PMID:26835549

  16. Foreign uranium supply

    International Nuclear Information System (INIS)

    Known foreign uranium resources are concentrated in a few countries. The resources of many countries are largely unassessed, but the known uranium countries appear to have the best potential for future expansion. Availability of supply from known resources will depend on resolution of national policies regarding uranium production, ownership and export, and actions of the mining industry. Foreign uranium demand projections have decreased markedly in the last few years, and currently planned and attainable production should be adequate through the 1980's. Longer term resources and supply outlook are still a major concern to both those planning electric supply systems based on converter reactors and those considering reprocessing and recycle of uranium and plutonium and development of breeder reactors. Work continues to clarify long-term supply in several countries and internationally, but more effort, and time, will be needed to clarify these issues

  17. Plutonium-enriched thermal fuel production experience in Belgium

    International Nuclear Information System (INIS)

    Taking into account the strategic aspects of nuclear energy such as availability and sufficiency of resources and independence of energy supply, most countries planning to use plutonium look mainly to its use in fast reactors. However, by recycling the recovered uranium and plutonium in light water reactors, the saving of the uranium that would otherwise be required could already be higher than 35%. Therefore, until fast reactors are introduced, for macro- or microeconomic reasons, the plutonium recycle option seems to be quite valuable for countries having the plutonium technology. In Belgium, Belgonucleaire has been developing the plutonium technology for more than 20 yr and has operated a mixed oxide fuel fabrication plant since 1973. The past ten years of plant operation have provided for many improvements and relevant new documented experiences establishing a basis for new modifications that will be beneficial to the intrinsic quality, overall safety, and economy of the fuel

  18. Standard practice for preparation and dissolution of plutonium materials for analysis

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This practice is a compilation of dissolution techniques for plutonium materials that are applicable to the test methods used for characterizing these materials. Dissolution treatments for the major plutonium materials assayed for plutonium or analyzed for other components are listed. Aliquants of the dissolved samples are dispensed on a weight basis when one of the analyses must be highly reliable, such as plutonium assay; otherwise they are dispensed on a volume basis. 1.2 The treatments, in order of presentation, are as follows: Procedure Title Section Dissolution of Plutonium Metal with Hydrochloric Acid 9.1 Dissolution of Plutonium Metal with Sulfuric Acid 9.2 Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed Oxide by the Sealed-Reflux Technique 9.3 Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed Oxides by Sodium Bisulfate Fusion 9.4 Dissolution of Uranium-Plutonium Mixed Oxides and Low-Fired Plutonium Oxide in Beakers 9.5 1.3 The values stated in SI units are to be re...

  19. Plutonium Story

    Science.gov (United States)

    Seaborg, G. T.

    1981-09-01

    The first nuclear synthesis and identification (i.e., the discovery) of the synthetic transuranium element plutonium (isotope /sup 238/Pu) and the demonstration of its fissionability with slow neutrons (isotope /sup 239/Pu) took place at the University of California, Berkeley, through the use of the 60-inch and 37-inch cyclotrons, in late 1940 and early 1941. This led to the development of industrial scale methods in secret work centered at the University of Chicago's Metallurgical Laboratory and the application of these methods to industrial scale production, at manufacturing plants in Tennessee and Washington, during the World War II years 1942 to 1945. The chemical properties of plutonium, needed to devise the procedures for its industrial scale production, were studied by tracer and ultramicrochemical methods during this period on an extraordinarily urgent basis. This work, and subsequent investigations on a worldwide basis, have made the properties of plutonium very well known. Its well studied electronic structure and chemical properties give it a very interesting position in the actinide series of inner transition elements.

  20. Plutonium story

    International Nuclear Information System (INIS)

    The first nuclear synthesis and identification (i.e., the discovery) of the synthetic transuranium element plutonium (isotope 238Pu) and the demonstration of its fissionability with slow neutrons (isotope 239Pu) took place at the University of California, Berkeley, through the use of the 60-inch and 37-inch cyclotrons, in late 1940 and early 1941. This led to the development of industrial scale methods in secret work centered at the University of Chicago's Metallurgical Laboratory and the application of these methods to industrial scale production, at manufacturing plants in Tennessee and Washington, during the World War II years 1942 to 1945. The chemical properties of plutonium, needed to devise the procedures for its industrial scale production, were studied by tracer and ultramicrochemical methods during this period on an extraordinarily urgent basis. This work, and subsequent investigations on a worldwide basis, have made the properties of plutonium very well known. Its well studied electronic structure and chemical properties give it a very interesting position in the actinide series of inner transition elements

  1. Storage of weapons-grade plutonium

    International Nuclear Information System (INIS)

    With the end of the cold war, the United States has started to reduce its nuclear weapons stockpiles and will place special nuclear material (plutonium and uranium) into storage. The existing plutonium storage facilities are designed for short-term storage or to support operation of adjacent processing facilities. The U.S. Department of Energy (DOE) is proposing to construct and operate a new plutonium storage facility as part of the reconfigured weapons complex, called Complex-21, to provide safe and secure long-term storage of plutonium materials. This facility will be required to meet new, more stringent requirements such as potential third-party inspection, enhanced safeguard and security requirements, and reduced personnel radiation exposure limits

  2. Plutonium, cesium, uranium and thorium series radionuclides in the Hudson River estuary and other environments. Annual technical progress report, 1 December 1984-30 November 1985

    International Nuclear Information System (INIS)

    We have measured radionuclide activities in a large number of sediment cores and suspended particle samples throughout the salinity range of the Hudson River estuary. Activities of 137Cs, 134Cs and 60Co determined by gamma spectrometry and /sup 239,240/Pu and 238Pu determined by alpha spectrometry indicate reasonably rapid accumulation rates in the sediments of marginal cove areas, and very rapid deposition in the harbor region adjacent to New York City, resulting in /sup 239,240/Pu accumulations there more than an order of magnitude greater than the fallout delivery rate. Fallout /sup 239,240/Pu moving downstream in the Hudson appears to be almost completely retained within the system by particle deposition, while more than 50% of the 137Cs derived from both reactor releases and fallout has been exported from the tidal Hudson to coastal waters. Some significant movement of dissolved plutonium into the estuary from the adjacent coastal waters may well be occurring. Depth profiles of radionuclides in Hudson sediments do not appear to be significantly altered by physical mixing processes in the sediment in areas accumulating particles at greater than 1 cm/y. Transport of fallout radionuclides from the drainage basin to the tidal Hudson appears to have decreased much faster than would be calculated from continuous removal from a well-mixed soil reservoir, indicating that sequestering of a substantial portion of the soil fallout burden has occurred in the watershed soils over the past two decades. Activities of 60Co in New York harbor sediments in 1984 averaged considerably higher than in 1979 and 1981, suggesting releases of this nuclide to the Hudson comparable to the first five years of reactor operations. 12 figs., 9 tabs

  3. Plutonium speciation affected by environmental bacteria

    International Nuclear Information System (INIS)

    Plutonium has no known biological utility, yet it has the potential to interact with bacterial cellular and extracellular structures that contain metal-binding groups, to interfere with the uptake and utilization of essential elements, and to alter cell metabolism. These interactions can transform plutonium from its most common forms, solid, mineral-adsorbed, or colloidal Pu(IV), to a variety of biogeochemical species that have much different physico-chemical properties. Organic acids that are extruded products of cell metabolism can solubilize plutonium and then enhance its environmental mobility, or in some cases facilitate plutonium transfer into cells. Phosphate- and carboxylate-rich polymers associated with cell walls can bind plutonium to form mobile biocolloids or Pu-laden biofilm/mineral solids. Bacterial membranes, proteins or redox agents can produce strongly reducing electrochemical zones and generate molecular Pu(III/IV) species or oxide particles. Alternatively, they can oxidize plutonium to form soluble Pu(V) or Pu(VI) complexes. This paper reviews research on plutonium-bacteria interactions and closely related studies on the biotransformation of uranium and other metals. (orig.)

  4. Study on the identification of organic and common anions in the pyrohydrolysis distillate of mixed uranium-plutonium carbide for the interference free determination of chlorine and fluorine by ion chromatography

    Energy Technology Data Exchange (ETDEWEB)

    Jeyakumar, Subbiah; Mishra, Vivekchandra Guruprasad; Das, Mrinal Kanti; Raut, Vaibhavi Vishwajeet; Sawant, Ramesh Mahadeo [Bhabha Atomic Research Centre, Mumbai (India). Radioanalytical Chemistry Div.; Ramakumar, Karanam Lakshminarayana [Bhabha Atomic Research Centre, Mumbai (India). Radiochemistry and Isotope Group

    2014-07-01

    Identification of various soluble organic acids formed during the pyrohydrolysis of uranium-plutonium mixed carbide [(U,Pu)C] was carried out using ion chromatography. This has significant importance as the soluble organic acids can cause severe interferences during the ion chromatography separation and determination of Cl{sup -} and F{sup -} in the pyrohydrolysis distillate of (U,Pu)C. Determination of Cl and F is important in the chemical quality control of nuclear materials as these two elements can cause corrosion and hence, their concentrations in all nuclear materials are restricted to certain specified values. Since the pyrohydrolysis distillates contain both inorganic and organic acid anions, for the sake of separating and identifying organic acid anions from the common inorganic anions, three independent isocratic elutions using varying concentrations of NaOH eluent were employed for the separation of weakly, moderately and strongly retained anions. It was observed that pyrohydrolysis of (U,Pu)C also produced soluble organic acids as in the case of nitric acid dissolution of UC. The present investigation revealed the presence of formic, acetic, propionic, butyric, oxalic acid anions in the pyrohydrolysis distillate of (U,Pu)C in trace or ultra-trace concentrations. The presence of each organic acid identified in the chromatogram was confirmed with spike addition as well as by separating them by capillary electrophoresis method. The presence of lower aliphatic acids viz. formic and acetic acids was reconfirmed by carrying out an independent separation with tetraborate eluent. It is suggested that nitric acid being formed during pyrohydrolysis could be responsible for the formation of organic acids. Based on the findings, an ion chromatography separation method has been proposed for the interference-free determination of chloride and fluoride in pyrohydrolysis distillate of (U,Pu)C. (orig.)

  5. Electrochemical oxidation of americium in nitric medium: study of reaction mechanisms; Oxydation de l'americium par voie electrochimique: etude des mecanismes

    Energy Technology Data Exchange (ETDEWEB)

    Picart, S.; Chartier, D.; Donnet, L.; Adnet, J.M. [CEA Valrho, (DCC/DRRV/SPHA), 30 - Marcoule (France)

    2000-07-01

    One alternative selected by the CEA for partitioning minor actinides from aqueous solutions containing fission products is the selective extraction of oxidized americium. This is the SESAME process (Selective Extraction and Separation of Americium by Means of Electrolysis) aimed to convert americium to oxidation state (VI) and then extract it with a specific extractant of high valences. This paper presents the study of the electrochemical oxidation of americium in nitric medium which represents an important stage of the process. The reaction can be divided into two main steps: oxidation of americium (III) to americium (IV), and then of americium (IV) to americium (VI). For the first oxidation step, a ligand L is needed to stabilize the intermediate species americium (IV) which disproportionates in its free form into americium (III) and (V). Phospho-tungstate or silico-tungstate are appropriate ligands because they are stable in concentrated nitric acid and show a great affinity for metallic cations at oxidation state (IV) (Table 1 lists the stability constants of americium (IV) complexes). The presence of the lacunary poly-anion lowers the potential of the americium (IV) / americium (Ill) redox pair (see Figure 5 for the diagram of the apparent formal potential of americium versus ligand concentration). This makes it thermodynamically possible to oxidize americium (III) into americium (IV) at the anode of an electrolyzer in nitric acid. For the second oxidation step, a strong oxidant redox mediator, like silver (II), is needed to convert complexed americium at oxidation state (IV) to oxidation state (V). The AmVL complex is then hydrolyzed to yield americyle (V) aqua ion. A spectroscopic Raman study with {sup 18}O labeled species showed that the oxygen atoms of the americyle moiety came from water. This indicates that water hydrolyzes the americium (V) complex to produce americyle (V) aqua ion, AmO{sub 2}{sup +}. This cation reacts with silver (Il) to give

  6. Standard test method for plutonium by Iron (II)/Chromium (VI) amperometric titration

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This test method covers the determination of plutonium in unirradiated nuclear-grade plutonium dioxide, uranium-plutonium mixed oxides with uranium (U)/plutonium (Pu) ratios up to 21, plutonium metal, and plutonium nitrate solutions. Optimum quantities of plutonium to measure are 7 to 15 mg. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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.

  7. Fabrication of uranium–americium mixed oxide pellet from microsphere precursors: Application of CRMP process

    Energy Technology Data Exchange (ETDEWEB)

    Remy, E. [Radiochemistry and Processes Department, CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze (France); Picart, S., E-mail: sebastien.picart@cea.fr [Radiochemistry and Processes Department, CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze (France); Delahaye, T. [Fuel Cycle Technology Department, CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze (France); Jobelin, I. [Radiochemistry and Processes Department, CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze (France); Lebreton, F.; Horlait, D. [Fuel Cycle Technology Department, CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze (France); Bisel, I. [Radiochemistry and Processes Department, CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze (France); Blanchart, P. [Heterogeneous Materials Research Group, Centre Européen de la Céramique, F-87068 Limoges (France); Ayral, A. [Institut Européen des Membranes, CNRS-ENSCM-UM2, CC47, University Montpellier 2, F-34095 Montpellier cedex 5 (France)

    2014-10-15

    Highlights: • Dust free process for (U,Am)O{sub 2} transmutation target fabrication. • Synthesis of U{sub 0.9}Am{sub 0.1}O{sub 2} mixed oxide microspheres from ion exchange resin. • Fabrication of dense U{sub 0.9}Am{sub 0.1}O{sub 2} pellet with 95% TD from mixed oxide microspheres. - Abstract: Mixed uranium–americium oxides are one of the materials envisaged for Americium Bearing Blankets dedicated to transmutation in fast neutron reactors. Recently, several processes have been developed in order to validate fabrication flowchart in terms of material specifications such as density and homogeneity but also to suggest simplifications for lowering industrial costs and hazards linked to dust generation of highly contaminating and irradiating compounds. This study deals with the application of an innovative route using mixed oxide microspheres obtained from metal loaded resin bead calcination, called Calcined Resin Microsphere Pelletization (CRMP). The synthesis of mixed oxide microsphere precursor of U{sub 0.9}Am{sub 0.1}O{sub 2±δ} is described as well as its characterisation. The use of this free-flowing precursor allows the pressing and sintering of one pellet of U{sub 0.9}Am{sub 0.1}O{sub 2±δ}. The ceramic obtained was characterised and results showed that its microstructure is dense and homogeneous and its density attains 95% of the theoretical density. This study validates the scientific feasibility of the CRMP process applied to the fabrication of uranium and americium-containing materials.

  8. Seaborg's Plutonium ?

    CERN Document Server

    Norman, Eric B; Telhami, Kristina E

    2014-01-01

    Passive x-ray and gamma-ray analysis was performed on UC Berkeley's EH&S Sample S338. The object was found to contain Pu-239 and no other radioactive isotopes. The mass of Pu-239 contained in this object was determined to be 2.0 +- 0.3 micrograms. These observations are consistent with the identification of this object being the 2.77-microgram plutonium oxide sample described by Glenn Seaborg and his collaborators as the first sample of Pu-239 that was large enough to be weighed.

  9. A Plutonium-Contaminated Wound, 1985, USA

    Energy Technology Data Exchange (ETDEWEB)

    Doran M. Christensen, DO, REAC/TS Associate Director and Staff Physician Eugene H. Carbaugh, CHP, Staff Scientist, Internal Dosimetry Manager, Pacific Northwest National Laboratory, Richland, Washington

    2012-02-02

    A hand injury occurred at a U.S. facility in 1985 involving a pointed shaft (similar to a meat thermometer) that a worker was using to remove scrap solid plutonium from a plastic bottle. The worker punctured his right index finger on the palm side at the metacarpal-phalangeal joint. The wound was not through-and- through, although it was deep. The puncture wound resulted in deposition of ~48 kBq of alpha activity from the weapons-grade plutonium mixture with a nominal 12 to 1 Pu-alpha to {sup 241}Am-alpha ratio. This case clearly showed that DTPA was very effective for decorporation of plutonium and americium. The case is a model for management of wounds contaminated with transuranics: (1) a team approach for dealing with all of the issues surrounding the incident, including the psychological, (2) early surgical intervention for foreign-body removal, (3) wound irrigation with DTPA solution, and (4) early and prolonged DTPA administration based upon bioassay and in vivo dosimetry.

  10. General consideration of effective plutonium utilization in future LWRs

    International Nuclear Information System (INIS)

    In this study, the potential of mixed oxide fueled light water reactors (MOX-LWRs), especially focusing on the high conversion type LWRs (HC-LWRs) such as FLWR are evaluated in terms of both economic aspect and effective use of plutonium. For economics consideration, relative economics positions of MOX-LWRs are clarified comparing the cost of electricity for uranium fueled LWRs (U-LWRs), MOX-LWRs and fast breeder reactors (FBRs) assuming future natural uranium price raise and variation of parameters such as construction cost and capacity factor. Also the economic superiority of MOX utilization against the uranium use is mentioned from the view point of plutonium credit concerning to the front-end fuel cycle cost. In terms of effective use of plutonium, comparative evaluations on plutonium mass balance in the cases of HC-LWR and high moderation type LWRs (HM-LWRs) taking into account plutonium quality (ratio of fissile to total plutonium) constraint in multiple recycling are performed as representative MOX utilization cases. Through this evaluation, the advantageous features of plutonium multiple recycling by HC-LWR are clarified. From all these results, merits of the introduction of HC-LWRs are discussed. (author)

  11. Plutonium Chemistry in the UREX+ Separation Processes

    Energy Technology Data Exchange (ETDEWEB)

    ALena Paulenova; George F. Vandegrift, III; Kenneth R. Czerwinski

    2009-10-01

    The project "Plutonium Chemistry in the UREX+ Separation Processes” is led by Dr. Alena Paulenova of Oregon State University under collaboration with Dr. George Vandegrift of ANL and Dr. Ken Czerwinski of the University of Nevada at Las Vegas. The objective of the project is to examine the chemical speciation of plutonium in UREX+ (uranium/tributylphosphate) extraction processes for advanced fuel technology. Researchers will analyze the change in speciation using existing thermodynamics and kinetic computer codes to examine the speciation of plutonium in aqueous and organic phases. They will examine the different oxidation states of plutonium to find the relative distribution between the aqueous and organic phases under various conditions such as different concentrations of nitric acid, total nitrates, or actinide ions. They will also utilize techniques such as X-ray absorbance spectroscopy and small-angle neutron scattering for determining plutonium and uranium speciation in all separation stages. The project started in April 2005 and is scheduled for completion in March 2008.

  12. The plutonium, it is natural; Le plutonium, c'est naturel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    Sine 1942, little time after the plutonium synthesis, two physician have proved the presence of plutonium in pitchblende from Canada. Some searches have confirmed its presence in others ores from Belgian Congo, Russia and Brazil. The results of analysis made on these materials show that the quantities of plutonium and uranium given in number of atoms or mass are in a ratio comprised between 3.6.10{sup -12} and 20.10{sup -12}. The results in micrograms by tons or Becquerels by ton are given in a table for pitchblende at 60% uranium concentration, for a usual ore at 3/1000 U and for the earth s crust (3 g/t U). (N.C.)

  13. Gamma-sources on the basis of metallic americium-241

    International Nuclear Information System (INIS)

    A batch of gamma-radiation sources has been manufactured from metallic americium-241 of isotopic purity, its activity varying from 0.08 to 0.93 GBq. The cores of the sources are high-purity americium metal condensate on a tantalum or stainless steel substrate prepared by thermal decomposition of 241Pu-241Am alloy in a high vacuum. 7 refs., 1 tab

  14. Fabrication and Irradiation Factors Influencing Plutonium Recycle Economics

    International Nuclear Information System (INIS)

    Two of the major factors that influence fuel cycle economics are irradiation performance and fuel fabrication costs. To recycle plutonium economically it is imperative that plutonium fuels are not limited to specific powers or burn-ups significantly below those attainable with uranium fuels. A major irradiation programme to compare the performance of plutonium fuels with uranium fuels is nearing completion. Fuel performance has been evaluated at powers high enough to cause central melting as well as with defected cladding. Irradiations of significant quantities of mixed uranium-plutonium oxide fuel elements have shown that fuels enriched with plutonium perform equally as well, and in some respects better, than fuels containing only uranium. The higher self-shielding obtained with plutonium-enriched fuels (compared with 235U-enriched fuels) results in an increase of 10 to 20% in power generation before melting occurs. For the PuO2 concentrations of interest in thermal reactors, fuel structural changes that occur in UO2 -PuO2 fuels during irradiation are very similar to those occurring in UO2 fuels. No difference in fission gas release rates and fuel swelling rates in UO2 and UC2 -PuO2 fuels has been determined. In-reactor temperatures at which sintering of physically mixed UO2 -PuO2 powders occurs causes homogenization and solid-solution formation by UO2 and PuO2 interdiffusion. Fabrication costs for both UO2 and UO2 -PuO2 processes have been studied in detail. These studies are based on actual fabrication experience extrapolated by comprehensive computer calculations to fuel requirements for a 1000 MW(e) pressurized water reactor. Fabrication costs have been analysed for plutonium and uranium-enriched fuel elements, fabricated by pelletizing and by vibrational compaction. The effect of throughput levels on fabrication costs has been studied extensively. (author)

  15. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Energy Technology Data Exchange (ETDEWEB)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  16. Determination of uranium isotopes in urine samples from radiation workers using 232U tracer, anion-exchange resin and alpha-spectrometry

    International Nuclear Information System (INIS)

    Bioassay technique is used for the estimation of actinides present in the body based on the excretion rate of body fluids. For occupational radiation workers urine assay is the preferred method for monitoring of chronic internal exposure. Determination of low concentrations of actinides such as plutonium, americium and uranium at low level of mBq in urine by alpha-spectrometry requires pre-concentration of large volumes of urine. This paper deals with standardization of analytical method for the determination of U-isotopes in urine samples using anion-exchange resin and 232U tracer for radiochemical recovery. The method involves oxidation of urine followed by co-precipitation of uranium along with calcium phosphate. Separation of U was carried out by Amberlite, IRA-400, anion-exchange resin. U-fraction was electrodeposited and activity estimated using tracer recovery by alpha-spectrometer. Eight routine urine samples of radiation workers were analyzed and consistent radiochemical tracer recovery was obtained in the range of 51% to 67% with a mean and standard deviation of 60% and 5.4%, respectively. (author)

  17. Electrochemical oxidation of americium in nitric medium: study of reaction mechanisms

    International Nuclear Information System (INIS)

    One alternative selected by the CEA for partitioning minor actinides from aqueous solutions containing fission products is the selective extraction of oxidized americium. This is the SESAME process (Selective Extraction and Separation of Americium by Means of Electrolysis) aimed to convert americium to oxidation state (VI) and then extract it with a specific extractant of high valences. This paper presents the study of the electrochemical oxidation of americium in nitric medium which represents an important stage of the process. The reaction can be divided into two main steps: oxidation of americium (III) to americium (IV), and then of americium (IV) to americium (VI). For the first oxidation step, a ligand L is needed to stabilize the intermediate species americium (IV) which disproportionates in its free form into americium (III) and (V). Phospho-tungstate or silico-tungstate are appropriate ligands because they are stable in concentrated nitric acid and show a great affinity for metallic cations at oxidation state (IV) (Table 1 lists the stability constants of americium (IV) complexes). The presence of the lacunary poly-anion lowers the potential of the americium (IV) / americium (Ill) redox pair (see Figure 5 for the diagram of the apparent formal potential of americium versus ligand concentration). This makes it thermodynamically possible to oxidize americium (III) into americium (IV) at the anode of an electrolyzer in nitric acid. For the second oxidation step, a strong oxidant redox mediator, like silver (II), is needed to convert complexed americium at oxidation state (IV) to oxidation state (V). The AmVL complex is then hydrolyzed to yield americyle (V) aqua ion. A spectroscopic Raman study with 18O labeled species showed that the oxygen atoms of the americyle moiety came from water. This indicates that water hydrolyzes the americium (V) complex to produce americyle (V) aqua ion, AmO2+. This cation reacts with silver (Il) to give americyle (VI) ion. Figure

  18. The future of plutonium - an overview

    International Nuclear Information System (INIS)

    Plutonium is the underpinning of the nuclear industry. Without it it is estimated that the fuel will run out not long after the turn of the century. With plutonium in fast breeders nuclear reactors can be operated for tens of thousands of years and the depleted uranium now available can be utilized The fuel cycle contemplated is similar to that of the light water reactor with some important differences at least partially related to the greater radioactivity of the resulting mixture of plutonium isotopes. The regulatory program does recognize the problems, including those of toxicity, safeguards and transportation. The concept of an integrated fuel cycle facility at a single location must be seriously considered. (author)

  19. Chemical Disposition of Plutonium in Hanford Site Tank Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Susan A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-05-07

    This report examines the chemical disposition of plutonium (Pu) in Hanford Site tank wastes, by itself and in its observed and potential interactions with the neutron absorbers aluminum (Al), cadmium (Cd), chromium (Cr), iron (Fe), manganese (Mn), nickel (Ni), and sodium (Na). Consideration also is given to the interactions of plutonium with uranium (U). No consideration of the disposition of uranium itself as an element with fissile isotopes is considered except tangentially with respect to its interaction as an absorber for plutonium. The report begins with a brief review of Hanford Site plutonium processes, examining the various means used to recover plutonium from irradiated fuel and from scrap, and also examines the intermediate processing of plutonium to prepare useful chemical forms. The paper provides an overview of Hanford tank defined-waste–type compositions and some calculations of the ratios of plutonium to absorber elements in these waste types and in individual waste analyses. These assessments are based on Hanford tank waste inventory data derived from separately published, expert assessments of tank disposal records, process flowsheets, and chemical/radiochemical analyses. This work also investigates the distribution and expected speciation of plutonium in tank waste solution and solid phases. For the solid phases, both pure plutonium compounds and plutonium interactions with absorber elements are considered. These assessments of plutonium chemistry are based largely on analyses of idealized or simulated tank waste or strongly alkaline systems. The very limited information available on plutonium behavior, disposition, and speciation in genuine tank waste also is discussed. The assessments show that plutonium coprecipitates strongly with chromium, iron, manganese and uranium absorbers. Plutonium’s chemical interactions with aluminum, nickel, and sodium are minimal to non-existent. Credit for neutronic interaction of plutonium with these absorbers

  20. Advanced fuels for plutonium management in pressurized water reactors

    International Nuclear Information System (INIS)

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h)-1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

  1. Theory of Antineutrino Monitoring of Burning MOX Plutonium Fuels

    CERN Document Server

    Hayes, A C; Nieto, Michael Martin; WIlson, W B

    2011-01-01

    This letter presents the physics and feasibility of reactor antineutrino monitoring to verify the burnup of plutonium loaded in the reactor as a Mixed Oxide (MOX) fuel. It examines the magnitude and temporal variation in the antineutrino signals expected for different MOX fuels, for the purposes of nuclear accountability and safeguards. The antineutrino signals from reactor-grade and weapons-grade MOX are shown to be distinct from those from burning low enriched uranium. Thus, antineutrino monitoring could be used to verify the destruction of plutonium in reactors, though verifying the grade of the plutonium being burned is found to be more challenging.

  2. Applicability of insoluble tannin to treatment of waste containing americium

    International Nuclear Information System (INIS)

    The applicability of insoluble tannin adsorbent to the treatment of aqueous waste contaminated with americium has been investigated. Insoluble tannin is considered highly applicable because it consists of only carbon, hydrogen and oxygen and so its volume can be easily reduced by incineration. This report describes measurements of the americium distribution coefficient in low concentration nitric acid. The americium distribution coefficients were found to decrease with increasing concentration of nitric acid and sodium nitrate, and with increasing temperature. At 25 C in 2.0 x 10-3 M HNO3, the distribution coefficient was found to be 2000 ml g-1. The adsorption capacity was determined by column experiments using europium as a simulant of americium, and found to be 7 x 10-3 mmol g-1-dried tannin in 0.01 M HNO3 at 25 C, which corresponds to approximately 1.7 mg-241Am/g-adsorbent(dried). The prospect of applying the adsorbent to the treatment of aqueous waste contaminated with americium appears promising. (orig.)

  3. Plutonium Finishing Plant

    Data.gov (United States)

    Federal Laboratory Consortium — The Plutonium Finishing Plant, also known as PFP, represented the end of the line (the final procedure) associated with plutonium production at Hanford.PFP was also...

  4. Measurement of plutonium in spent nuclear fuel by self-induced x-ray fluorescence

    Energy Technology Data Exchange (ETDEWEB)

    Hoover, Andrew S [Los Alamos National Laboratory; Rudy, Cliff R [Los Alamos National Laboratory; Tobin, Steve J [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Stafford, A [TEXAS A& M; Strohmeyer, D [TEXAS A& M; Saavadra, S [ORNL

    2009-01-01

    Direct measurement of the plutonium content in spent nuclear fuel is a challenging problem in non-destructive assay. The very high gamma-ray flux from fission product isotopes overwhelms the weaker gamma-ray emissions from plutonium and uranium, making passive gamma-ray measurements impossible. However, the intense fission product radiation is effective at exciting plutonium and uranium atoms, resulting in subsequent fluorescence X-ray emission. K-shell X-rays in the 100 keV energy range can escape the fuel and cladding, providing a direct signal from uranium and plutonium that can be measured with a standard germanium detector. The measured plutonium to uranium elemental ratio can be used to compute the plutonium content of the fuel. The technique can potentially provide a passive, non-destructive assay tool for determining plutonium content in spent fuel. In this paper, we discuss recent non-destructive measurements of plutonium X-ray fluorescence (XRF) signatures from pressurized water reactor spent fuel rods. We also discuss how emerging new technologies, like very high energy resolution microcalorimeter detectors, might be applied to XRF measurements.

  5. PLUTONIUM-THORIUM ALLOYS

    Science.gov (United States)

    Schonfeld, F.W.

    1959-09-15

    New plutonium-base binary alloys useful as liquid reactor fuel are described. The alloys consist of 50 to 98 at.% thorium with the remainder plutonium. The stated advantages of these alloys over unalloyed plutonium for reactor fuel use are easy fabrication, phase stability, and the accompanying advantuge of providing a means for converting Th/sup 232/ into U/sup 233/.

  6. PLUTONIUM CLEANING PROCESS

    Science.gov (United States)

    Kolodney, M.

    1959-12-01

    A method is described for rapidly removing iron, nickel, and zinc coatings from plutonium objects while simultaneously rendering the plutonium object passive. The method consists of immersing the coated plutonium object in an aqueous acid solution containing a substantial concentration of nitrate ions, such as fuming nitric acid.

  7. Measurements of plutonium residues from recovery processes

    International Nuclear Information System (INIS)

    Conventional methods of nondestructive assay (NDA) have accurately assayed the plutonium content of many forms of relatively pure and homogeneous bulk items. However, physical and chemical heterogeneities and the high and variable impurity levels of many categories of processing scrap bias the conventional NDA results. The materials also present a significant challenge to the assignment of reference values to process materials for purposes of evaluating the NDA methods. A recent study using impure, heterogeneous, pyrochemical residues from americium molten salt extraction (MSE) has been aimed at evaluating NDA assay methods based on conventional gamma-ray and neutron measurement techniques and enhanced with analyses designed to address the problems of heterogeneities and impurities. The study included a significant effort to obtain reference values for the MSE spent salts used in the study. Two of the improved NDA techniques, suitable for in-line assay of plutonium in bulk, show promise for timely in-process assays for one of the most difficult pyrochemical residues generated as well as for other impure heterogeneous scrap categories. 12 refs., 4 figs., 5 tabs

  8. Nuclear power fleets and uranium resources recovered from phosphates

    International Nuclear Information System (INIS)

    Current light water reactors (LWR) burn fissile uranium, whereas some future reactors, as Sodium fast reactors (SFR) will be capable of recycling their own plutonium and already-extracted depleted uranium. This makes them a feasible solution for the sustainable development of nuclear energy. Nonetheless, a sufficient quantity of plutonium is needed to start up an SFR, with the plutonium already being produced in light water reactors. The availability of natural uranium therefore has a direct impact on the capacity of the reactors (both LWR and SFR) that we can build. It is therefore important to have an accurate estimate of the available uranium resources in order to plan for the world's future nuclear reactor fleet. This paper discusses the correspondence between the resources (uranium and plutonium) and the nuclear power demand. Sodium fast reactors will be built in line with the availability of plutonium, including fast breeders when necessary. Different assumptions on the global uranium resources are taken into consideration. The largely quoted estimate of 22 Mt of uranium recovered for phosphate rocks can be seriously downscaled. Based on our current knowledge of phosphate resources, 4 Mt of recoverable uranium already seems to be an upper bound value. The impact of the downscaled estimate on the deployment of a nuclear fleet is assessed accordingly. (authors)

  9. Selected measurement data for plutonium and uranium

    International Nuclear Information System (INIS)

    This handbook is designed primarily for use by inspectors when verifying the nuclear material contents of samples by nondestructive means. It contains discussions of various properties and the applicability of the measurement of these properties toward verification of specific samples. Because the thrust of this handbook is with properties rather than techniques or instruments, portions of it may be useful to other analysts as a reference

  10. The treatment of liquid radioactive waste containing Americium by using a cation exchange method

    International Nuclear Information System (INIS)

    A research in the treatment of a liquid radioactive waste containing americium has been done. The liquid radioactive waste used in this research was standard solution of U dan Ce with the initial activity of 100 ppm. The experimental investigation is aimed at a study of the effects of the waste pH, the column dimension of IR-120 cation exchanger which is expressed as L/D, the flow rate of a liquid waste and the influence of thiocyanate as a complex agent against the efficiency of a decontamination for uranium and cerium element. The experiment was done by passing downward the feed of uranium and cerium solution into an IR-120 type of cation exchanger with the L/D of 11.37. From the experimental parameters done in this research where the influence of waste pH was varied from 3 - 8, the geometric column (L/D) 11.37, the liquid flow rate was from 2.5 - 10 ml/m and the thiocyanate concentration was between 100 ppm-500 ppm can be concluded that the optimum operational condition for the ion exchange achieved were the waste pH for uranium = 4 and the waste pH for cerium = 6, the flow rate = 2.5 ml/men. From the given maximum value of DF for uranium = 24 (DE = 95.83%) and of DF for cerium = 40 (DE = 97.5%), it can also be concluded that this investigation is to be continued in order that the greater value of DF/DE can be achieved

  11. The use of carbohydrazide for plutonium concentration stripping in separator with inert packing

    International Nuclear Information System (INIS)

    For the purpose of removing plutonium from uranium- plutonium extract it is proposed to employ concentration stripping process with the use of separator and a new reducing reagent: Carbohydrazide CO(N2H3)2. Using plutonium stripping from solution simulating the composition of extract of spent nuclear fuel from VVER-1000 reactor (without γ-emitting isotopes), with O: A ratio of = 28, a product solution was obtained containing 17.8 g/l of plutonium, 29.2 g/l of uranium and more than 1 g/l of technetium. The experiment on real spent fuel from VVER-1000 with burn-up of more than 50 GW*d/t of uranium after 17 year exposure, performed in the shielded box of FSUE 'MCP', confirmed the effectiveness and feasibility of the proposed process. Through concentration stripping (O:A = 20), a plutonium product solution was obtained with a part of uranium with the following composition: [U] = 150 g/l; [Pu] = 23,5 g/l; [Np] = 1,7 g/l, [Tc] = 1.5 g/l; gamma exposure rate - 0,022 mR/s*l. Direct extraction of plutonium in this operation was 95.3%, the rest of plutonium is refluxing to the preceding stage of the extraction cycle. A process flow diagram with organization of plutonium recycling is proposed, allowing for its complete removal into a single stream. Carbohydrazide is an effective reducing agent of plutonium (IV), ensuring the stability of uranium-plutonium separation process. (authors)

  12. Russian plutonium policy

    Energy Technology Data Exchange (ETDEWEB)

    Ponomarev-Stepnoi, N.; Tsourikov, D. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation)

    1997-10-01

    This paper is intended to provide more detail on the main features of Russian strategy of utilization of both the civilian and weapons-grade plutonium. At present, the Russian Federation has a large stock of plutonium and at the same time some scientific, technological and industrial experience in the utilization of plutonium, in particular in fast reactors. The key elements of Russian plutonium policy are the interim secure storage and plutonium disposition in nuclear reactors. The disposition options being discussed are the following: BN-type reactors, VVERs, and HTGR. It is shown that the utilization of weapons-grade plutonium, for a number of reasons, should begin using the reactors currently in operation. The importance of broad international cooperation for a safe and effective management of weapons plutonium designated as no longer required for defense purposes has been stressed. (orig.)

  13. The proliferation potential of neptunium and americium

    Energy Technology Data Exchange (ETDEWEB)

    An, J. S.; Shin, J. S.; Kim, J. S.; Kwack, E. H.; Kim, B. K

    2000-05-01

    It is recognized that some trans-uranic elements other than plutonium, in particular Np and Am, if will be available in sufficient quantities, could be used for nuclear explosive devices. The spent fuel has been accumulating in number of nuclear power plant and operation of large scale commercial reprocessing plants. However, these materials are not covered by the definition of special fissionable material in the Agency Statute. At the time when the Statute was adopted, the availability of meaningful quantities of separated Np and Am was remote and they were not included in the definition of special fissionable material. Then, IAEA Board decided a measure for control of Np and Am on September 1999. This report contains the control method and the characteristic of Np and Am for using domestic nuclear industries, and it can be useful for understanding how to report and account of Np and Am. (author)

  14. Technical report for generic site add-on facility for plutonium polishing

    International Nuclear Information System (INIS)

    The purpose of this report is to provide environmental data and reference process information associated with incorporating plutonium polishing steps (dissolution, impurity removal, and conversion to oxide powder) into the genetic-site Mixed-Oxide Fuel Fabrication Facility (MOXFF). The incorporation of the plutonium polishing steps will enable the removal of undesirable impurities, such as gallium and americium, known to be associated with the plutonium. Moreover, unanticipated impurities can be removed, including those that may be contained in (1) poorly characterized feed materials, (2) corrosion products added from processing equipment, and (3) miscellaneous materials contained in scrap recycle streams. These impurities will be removed to the extent necessary to meet plutonium product purity specifications for MOX fuels. Incorporation of the plutonium polishing steps will mean that the Pit Disassembly and Conversion Facility (PDCF) will need to produce a plutonium product that can be dissolved at the MOXFF in nitric acid at a suitable rate (sufficient to meet overall production requirements) with the minimal usage of hydrofluoric acid, and its complexing agent, aluminum nitrate. This function will require that if the PDCF product is plutonium oxide powder, that powder must be produced, stored, and shipped without exceeding a temperature of 600 C

  15. Studies of a liquid anode for plutonium electrorefining

    International Nuclear Information System (INIS)

    They are developing a solvent anode as an alternate method for producing plutonium metal of high purity by an electrorefining process. The goals are to produce metal of 99.98% purity with an anode residue containing less than 2% of the plutonium in the feed material. If they are successful, they will design and demonstrate a system utilizing semi-continuous and remotely controlled operations. Establishing a solvent anode method should lead to improved yields and a substantial reduction in the amount of residues generated by the electrorefining process. The new method should be a viable pyrochemical technique for recovering both plutonium and uranium from spent reactor fuel. Initially, the anode consists of a tantalum rod immersed in a pool of liquid cadmium at 7400C. Impure plutonium or a solid anode residue is in contact with the cadmium. As current passes through the anode, plutonium in the cadmium is oxidized and transfers into the molten salt as tripositive plutonium. More plutonium dissolves into the cadmium and the oxidation continues. The tri-positive plutonium is carried through the salt to the cathode, where it is reduced to pure, liquid metal. This metal, which is heavier than the salt, drips from the cathode into an annulus between the anode and cathode compartments and forms a product ring

  16. The first law of uranium dynamics

    International Nuclear Information System (INIS)

    An embargo on the export of Australia's vast uranium mineral resources has been in force for several years. Pressures to lift the embargo are mounting. A Royal Commission has enquired into all aspects of uranium mining and exports. This paper, modified to some extent forms part of a submission to this enquiry. It is suggested that nuclear power be accepted as an interim solution to the world's energy problems. The plutonium fuelled breeder technology as the ultimate solution is rejected. Meanwhile the striking of a $750 (Australian) export levy per tonne of U3O8 will provide the funds for research and development in the use of renewable energy sources. Restrictive conditions imposed on the scale of uranium minerals and their beneficiated derivatives include guarantees against the separation of plutonium and the development and use of a plutonium technology. (author)

  17. Weapons-grade plutonium dispositioning. Volume 2: Comparison of plutonium disposition options

    Energy Technology Data Exchange (ETDEWEB)

    Brownson, D.A.; Hanson, D.J.; Blackman, H.S. [and others

    1993-06-01

    The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate disposition options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) offered to assist the NAS in this evaluation by investigating the technical aspects of the disposition options and their capability for achieving plutonium annihilation levels greater than 90%. This report was prepared for the NAS to document the gathered information and results from the requested option evaluations. Evaluations were performed for 12 plutonium disposition options involving five reactor and one accelerator-based systems. Each option was evaluated in four technical areas: (1) fuel status, (2) reactor or accelerator-based system status, (3) waste-processing status, and (4) waste disposal status. Based on these evaluations, each concept was rated on its operational capability and time to deployment. A third rating category of option costs could not be performed because of the unavailability of adequate information from the concept sponsors. The four options achieving the highest rating, in alphabetical order, are the Advanced Light Water Reactor with plutonium-based ternary fuel, the Advanced Liquid Metal Reactor with plutonium-based fuel, the Advanced Liquid Metal Reactor with uranium-plutonium-based fuel, and the Modular High Temperature Gas-Cooled Reactor with plutonium-based fuel. Of these four options, the Advanced Light Water Reactor and the Modular High Temperature Gas-Cooled Reactor do not propose reprocessing of their irradiated fuel. Time constraints and lack of detailed information did not allow for any further ratings among these four options. The INEL recommends these four options be investigated further to determine the optimum reactor design for plutonium disposition.

  18. Weapons-grade plutonium dispositioning. Volume 2: Comparison of plutonium disposition options

    International Nuclear Information System (INIS)

    The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate disposition options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) offered to assist the NAS in this evaluation by investigating the technical aspects of the disposition options and their capability for achieving plutonium annihilation levels greater than 90%. This report was prepared for the NAS to document the gathered information and results from the requested option evaluations. Evaluations were performed for 12 plutonium disposition options involving five reactor and one accelerator-based systems. Each option was evaluated in four technical areas: (1) fuel status, (2) reactor or accelerator-based system status, (3) waste-processing status, and (4) waste disposal status. Based on these evaluations, each concept was rated on its operational capability and time to deployment. A third rating category of option costs could not be performed because of the unavailability of adequate information from the concept sponsors. The four options achieving the highest rating, in alphabetical order, are the Advanced Light Water Reactor with plutonium-based ternary fuel, the Advanced Liquid Metal Reactor with plutonium-based fuel, the Advanced Liquid Metal Reactor with uranium-plutonium-based fuel, and the Modular High Temperature Gas-Cooled Reactor with plutonium-based fuel. Of these four options, the Advanced Light Water Reactor and the Modular High Temperature Gas-Cooled Reactor do not propose reprocessing of their irradiated fuel. Time constraints and lack of detailed information did not allow for any further ratings among these four options. The INEL recommends these four options be investigated further to determine the optimum reactor design for plutonium disposition

  19. Uses of Plutonium Fuel in Pressure-Tube-Type, Heavy-Water-Moderated Thermal Reactors

    International Nuclear Information System (INIS)

    In 1962, a feasibility study was begun in the JAERI on the uses of various nuclear fuels for pressure-tube-type, heavy-water-moderated thermal reactors. This study began with analysis of the use of uranium in heavy-water-moderated thermal reactors such as the CANDU-PHW, CANDU-BLW, SGHW, EL-4, and Ref. 15, D and E lattices, which is designed in the JAERI, from the standpoint of the core design. Then, the ways of using plutonium fuel in the same types were investigated using WATCHTOWER, FLARE and VENUS codes, including: (1) direct substitution of the plutonium from light-water reactors or Magnox reactors, (2) recycle use of the plutonium from heavy-water-moderated reactors, (3) plutonium self-sustaining cycle, and (4) plutonium phoenix fuel. The following conclusions are reported: (1) In the direct substitution of plutonium, somewhat depleted plutonium is more suitable for core design than the plutonium from Magnox reactors or light-water reactors, because the increase in the initial reactivity due to large plutonium absorption cross-section must be prevented. (2) In the plutonium self-sustaining cycle, the fuel burn-up of about 15 000 ∼20000 MWd/t would be expected from natural uranium, and the positive void reactivity which always occurs in the uraniumloaded SGHW or CANDU-BLW lattices is greatly reduced, the latter property giving some margin to bum-out heat flux. (3) It may be concluded from the fuel cycle analysis that the plutonium self-sustaining cycle is equivalent to using slightly enriched uranium (about 1.0 at.%). It may be concluded that the use of plutonium in heavy-water-moderated reactors is technologically feasible and economically advantageous. (author)

  20. 1976 Hanford americium exposure incident: overview and perspective

    International Nuclear Information System (INIS)

    Salient features of the 1976 Hanford americium exposure incident are discussed. Comparisons are made with previous human and animal exposure data, and conclusions drawn relative to the injured workman, to health physics practices, and to the adequacy of current exposure limits

  1. Application of Prompt Gamma-Ray Analysis to Identify Electrorefining Salt-Bearing Plutonium Oxide at the Plutonium Finishing Plant

    International Nuclear Information System (INIS)

    Prompt gamma-ray analysis is being implemented at the Plutonium Finishing Plant (PFP) to screen impure plutonium oxide inventory items, received in the mid-1980s from the Rocky Flats Plant, for the presence of sodium chloride and potassium chloride salts from the electrorefining process. A large fraction of these items are suspected to contain electrorefining salts. Because the salts evaporate at the=950C stabilization temperature mandated for long-term storage under the U.S. Department of Energy plutonium oxide stabilization and storage criteria to plug and corrode process equipment, items found to have these salts qualify for thermal stabilization at 750C. The prompt gamma ray energies characteristic of sodium, potassium, chlorine, and other low atomic weight elements arise from the interaction the light elements with alpha radiation from plutonium and americium radioactive decay. High-resolution gamma ray spectrometers designed to detect energies up to ∼4.5 MeV are used to gather the high-energy prompt gamma spectra.Observation of the presence of the high-energy gamma peaks representing the natural chlorine-35, sodium-23, and potassium-39 isotopes and the sodium-to-chlorine peak area ratios in the range for plutonium oxide materials known to contain the electrorefining salts give the evidence needed to identify plutonium oxide materials at the PFP that qualify for the lower-temperature processing. Conversely, the absence of these telltale signals in the prompt gamma analysis provides evidence that the materials do not contain the electrorefining salts. Furthermore, based on calibrations using known assayed items, semiquantitative measurement of the quantity of chlorine present in materials containing electrorefining salt also can be performed by using the count rates observed for the chlorine peak, the plutonium quantity present in the measured item, and the plutonium- and chlorine-specific response of the gamma detection system. The origin and characteristics

  2. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    International Nuclear Information System (INIS)

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium

  3. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Bronson, M.C.

    1997-10-01

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium.

  4. Imperatives for using plutonium in commercial power reactors

    International Nuclear Information System (INIS)

    The use of reprocessed or newly produced plutonium as a fissile fuel in commercial nuclear reactors in the US has been actively suppressed by the current US Administration. Yet, many other advanced nations have already adopted mixed oxide fuels which are manufactured from a mixture of plutonium and natural uranium compounds. These nations have successfully proven the use of such nuclear fuel in their commercial power reactors for many years. The full consequence of the restrictive nuclear policy in the US will greatly limit the lifetime of the nuclear fuel resources in the US from a nominal potential of 100 centuries or more of potential energy supply to about 50 years or less at economical prices for uranium. This paper addresses both the imperatives and the potential and the perceived hazards of plutonium utilization and examines the consequences of government policy regarding utilization of nuclear power

  5. Plutonium fuel an assessment. Report by an expert group

    International Nuclear Information System (INIS)

    Since the 1950s, plutonium used in fast reactors has been seen as the key to unlocking the vast energy resources contained in the world's uranium reserves. However, the slowing down in projected installation rates of nuclear reactors, combined with discovery of additional uranium, have led to a postponement of the point in time when fast reactors will make large demands on plutonium supplies. There are several options concerning its use or storage in the meantime. This report sets out the facts and current views about plutonium and its civil use, both at present and in the medium term. It explains the factors influencing the choice of fuel options and illustrates how economic and logistic assessments of the alternatives can be undertaken

  6. Plutonium measurements in the environment of the nuclear site of Marcoule

    International Nuclear Information System (INIS)

    A contamination of ground higher than the national average has been measured on a 25 km2 surface around the site. This corresponds to: 3 giga-becquerels of plutonium distributed on this surface that represents a surface activity of 120 Bq/m2. In France this average activity of the plutonium is about 50 Bq/m2; 0.6 giga-becquerels of americium 241, distributed on this same surface, that is to say a surface activity about 24 Bq/m2. In France this average surface activity (of Am 241) is in the order of 20 Bq/m2. The low quantities of plutonium cannot induce to a significant exposure for the populations of concerned communes (the calculated doses are inferior to 0.001 mSv by year). (N.C.)

  7. Reduction Rates for Higher Americium Oxidation States in Nitric Acid

    Energy Technology Data Exchange (ETDEWEB)

    Grimes, Travis Shane [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mincher, Bruce Jay [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schmitt, Nicholas C [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-30

    The stability of hexavalent americium was measured using multiple americium concentrations and nitric acid concentrations after contact with the strong oxidant sodium bismuthate. Contrary to our hypotheses Am(VI) was not reduced faster at higher americium concentrations, and the reduction was only zero-order at short time scales. Attempts to model the reduction kinetics using zero order kinetic models showed Am(VI) reduction in nitric acid is more complex than the autoreduction processes reported by others in perchloric acid. The classical zero-order reduction of Am(VI) was found here only for short times on the order of a few hours. We did show that the rate of Am(V) production was less than the rate of Am(VI) reduction, indicating that some Am(VI) undergoes two electron-reduction to Am(IV). We also monitored the Am(VI) reduction in contact with the organic diluent dodecane. A direct comparison of these results with those in the absence of the organic diluent showed the reduction rates for Am(VI) were not statistically different for both systems. Additional americium oxidations conducted in the presence of Ce(IV)/Ce(III) ions showed that Am(VI) is reduced without the typical growth of Am(V) observed in the systems sans Ce ion. This was an interesting result which suggests a potential new reduction/oxidation pathway for Am in the presence of Ce; however, these results were very preliminary, and will require additional experiments to understand the mechanism by which this occurs. Overall, these studies have shown that hexavalent americium is fundamentally stable enough in nitric acid to run a separations process. However, the complicated nature of the reduction pathways based on the system components is far from being rigorously understood.

  8. Americium and plutonium release behavior from irradiated mixed oxide fuel during heating

    Science.gov (United States)

    Sato, I.; Suto, M.; Miwa, S.; Hirosawa, T.; Koyama, S.

    2013-06-01

    The release behavior of Pu and Am was investigated under the reducing atmosphere expected in sodium cooled fast reactor severe accidents. Irradiated Pu and U mixed oxide fuels were heated at maximum temperatures of 2773 K and 3273 K. EPMA, γ-ray spectrometry and α-ray spectrometry for released and residual materials revealed that Pu and Am can be released more easily than U under the reducing atmosphere. The respective release rate coefficients for Pu and Am were obtained as 3.11 × 10-4 min-1 and 1.60 × 10-4 min-1 at 2773 K under the reducing atmosphere with oxygen partial pressure less than 0.02 Pa. Results of thermochemical calculations indicated that the main released chemical forms would likely be PuO for Pu and Am for Am under quite low oxygen partial pressure.

  9. Americium and plutonium separation by extraction chromatography for determination by accelerator mass spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Kazi, Zakir H. [Department of Earth Science, University of Ottawa, 140 Louis Pasteur Avenue, Ottawa K1N 6N5 (Canada); Cornett, Jack R., E-mail: jack.cornett@uottawa.ca [Department of Earth Science, University of Ottawa, 140 Louis Pasteur Avenue, Ottawa K1N 6N5 (Canada); Zhao, Xaiolei; Kieser, Liam [Department of Physics, University of Ottawa, 140 Louis Pasteur Avenue, Ottawa K1N 6N5 (Canada)

    2014-06-01

    Highlights: • Am and Pu were adsorbed and separated using a single extraction chromatography DGA column. • Pu was eluted from the column completely using on-column reduction of Pu(IV) to Pu(III). • ²⁴¹Am and 239,240Pu measurements by accelerator mass spectrometry (AMS) agree with the certified values in two SRMs. Abstract: A simple method was developed to separate Pu and Am using single column extraction chromatography employing N,N,N',N'-tetra-n-octyldiglycolamide (DGA) resin. Isotope dilution measurements of Am and Pu were performed using accelerator mass spectrometry (AMS) and alpha spectrometry. For maximum adsorption Pu was stabilized in the tetra valent oxidation state in 8 M HNO₃ with 0.05 M NaNO₂ before loading the sample onto the resin. Am(III) was adsorbed also onto the resin from concentrated HNO₃, and desorbed with 0.1 M HCl while keeping the Pu adsorbed. The on-column reduction of Pu(IV) to Pu(III) with 0.02 M TiCl₃ facilitated the complete desorption of Pu. Interferences (e.g. Ca²⁺, Fe³⁺) were washed off from the resin bed with excess HNO₃. Using NdF₃, micro-precipitates of the separated isotopes were prepared for analysis by both AMS and alpha spectrometry. The recovery was 97.7 ± 5.3% and 95.5 ± 4.6% for ²⁴¹Am and ²⁴²Pu respectively in reagents without a matrix. The recoveries of the same isotopes were 99.1 ± 6.0 and 96.8 ± 5.3% respectively in garden soil. The robustness of the method was validated using certified reference materials (IAEA 384 and IAEA 385). The measurements agree with the certified values over a range of about 1–100 Bq kg⁻¹. The single column separation of Pu and Am saves reagents, separation time, and cost.

  10. Plutonium and americium in air on the coasts of the Irish Sea

    International Nuclear Information System (INIS)

    Measurements of artificial radionuclides in air, deposition and seawater were carried out at coastal and inland locations in Cumbria, south-west Scotland, Northern Ireland and North Wales. The measurements were undertaken to determine the distribution of artificial radionuclides in the coastal environment attributable to discharges from Sellafield and to investigate the mechanisms responsible for the increased radionuclide concentration in the marine aerosol and their subsequent transfer to land. A number of different sampling methods were used and the contribution each method could make to the understanding of the marine aerosol was assessed. Measurements of the size distribution of the marine aerosol was also undertaken. The measurements demonstrated that the actinide concentrations measured in the air and deposition at coastal locations could be related to near shore sea water concentrations, but that no direct proportionality existed. It was also determined that a significant proportion of the material transferred from sea to land was associated with larger particulate present in the marine aerosol during conditions of above average onshore winds. (author)

  11. Concordant plutonium-241-americium-241 dating of environmental samples: results from forest fire ash

    Energy Technology Data Exchange (ETDEWEB)

    Goldstein, Steven J [Los Alamos National Laboratory; Oldham, Warren J [Los Alamos National Laboratory; Murrell, Michael T [Los Alamos National Laboratory; Katzman, Danny [Los Alamos National Laboratory

    2010-12-07

    We have measured the Pu, {sup 237}Np, {sup 241}Am, and {sup 151}Sm isotopic systematics for a set of forest fire ash samples from various locations in the western U.S. including Montana, Wyoming, Idaho, and New Mexico. The goal of this study is to develop a concordant {sup 241}Pu (t{sub 1/2} = 14.4 y)-{sup 241}Am dating method for environmental collections. Environmental samples often contain mixtures of components including global fallout. There are a number of approaches for subtracting the global fallout component for such samples. One approach is to use {sup 242}/{sup 239}Pu as a normalizing isotope ratio in a three-isotope plot, where this ratio for the nonglobal fallout component can be estimated or assumed to be small. This study investigates a new, complementary method of normalization using the long-lived fission product, {sup 151}Sm (t{sub 1/2} = 90 y). We find that forest fire ash concentrates actinides and fission products with {approx}1E10 atoms {sup 239}Pu/g and {approx}1E8 atoms {sup 151}Sm/g, allowing us to measure these nuclides by mass spectrometric (MIC-TIMS) and radiometric (liquid scintillation counting) methods. The forest fire ash samples are characterized by a western U.S. regional isotopic signature representing varying mixtures of global fallout with a local component from atmospheric testing of nuclear weapons at the Nevada Test Site (NTS). Our results also show that {sup 151}Sm is well correlated with the Pu nuclides in the forest fire ash, suggesting that these nuclides have similar geochemical behavior in the environment. Results of this correlation indicate that the {sup 151}Sm/{sup 239}Pu atom ratio for global fallout is {approx}0.164, in agreement with an independent estimate of 0.165 based on {sup 137}Cs fission yields for atmospheric weapons tests at the NTS. {sup 241}Pu-{sup 241}Am dating of the non-global fallout component in the forest fire ash samples yield ages in the late 1950's-early 1960's, consistent with a peak in NTS weapons testing at that time. The age results for this component are in agreement using both {sup 242}Pu and {sup 151}Sm normalizations, although the errors for the {sup 151}Sm correction are currently larger due to the greater uncertainty of their measurements. Additional efforts to develop a concordant {sup 241}Pu-{sup 241}Am dating method for environmental collections are underway with emphasis on soil cores.

  12. Americium and plutonium release behavior from irradiated mixed oxide fuel during heating

    International Nuclear Information System (INIS)

    The release behavior of Pu and Am was investigated under the reducing atmosphere expected in sodium cooled fast reactor severe accidents. Irradiated Pu and U mixed oxide fuels were heated at maximum temperatures of 2773 K and 3273 K. EPMA, γ-ray spectrometry and α-ray spectrometry for released and residual materials revealed that Pu and Am can be released more easily than U under the reducing atmosphere. The respective release rate coefficients for Pu and Am were obtained as 3.11 × 10−4 min−1 and 1.60 × 10−4 min−1 at 2773 K under the reducing atmosphere with oxygen partial pressure less than 0.02 Pa. Results of thermochemical calculations indicated that the main released chemical forms would likely be PuO for Pu and Am for Am under quite low oxygen partial pressure

  13. Americium and plutonium separation by extraction chromatography for determination by accelerator mass spectrometry.

    Science.gov (United States)

    Kazi, Zakir H; Cornett, Jack R; Zhao, Xaiolei; Kieser, Liam

    2014-06-01

    A simple method was developed to separate Pu and Am using single column extraction chromatography employing N,N,N',N'-tetra-n-octyldiglycolamide (DGA) resin. Isotope dilution measurements of Am and Pu were performed using accelerator mass spectrometry (AMS) and alpha spectrometry. For maximum adsorption Pu was stabilized in the tetra valent oxidation state in 8M HNO3 with 0.05 M NaNO2 before loading the sample onto the resin. Am(III) was adsorbed also onto the resin from concentrated HNO3, and desorbed with 0.1 M HCl while keeping the Pu adsorbed. The on-column reduction of Pu(IV) to Pu(III) with 0.02 M TiCl3 facilitated the complete desorption of Pu. Interferences (e.g. Ca(2+), Fe(3+)) were washed off from the resin bed with excess HNO3. Using NdF3, micro-precipitates of the separated isotopes were prepared for analysis by both AMS and alpha spectrometry. The recovery was 97.7±5.3% and 95.5±4.6% for (241)Am and (242)Pu respectively in reagents without a matrix. The recoveries of the same isotopes were 99.1±6.0 and 96.8±5.3% respectively in garden soil. The robustness of the method was validated using certified reference materials (IAEA 384 and IAEA 385). The measurements agree with the certified values over a range of about 1-100 Bq kg(-1). The single column separation of Pu and Am saves reagents, separation time, and cost. PMID:24856406

  14. Subsurface Behavior of Plutonium and Americium at Non-Hanford Sites and Relevance to Hanford

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Riley, Robert G.

    2008-02-01

    Seven sites where Pu release to the environment has raised significant environmental concerns have been reviewed. A summary of the most significant hydrologic and geochemical features, contaminant release events and transport processes relevant to Pu migration at the seven sites is presented.

  15. Properties of plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Jin Su; Yoon, Hwan Ki; Min, Kyung Sik; Kim, Hyun Tae; Ahn, Jong Sung; Kwag, Eon Ho; Ryu, Keon Joong [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of)

    1996-03-01

    Plutonium has unique chemical and physical properties. Its uniqueness in use has led to rare publications, in Korea. This report covers physical aspects of phase change of metal plutonium, mechanical properties, thermal conductivity, etc, chemical aspects of corrosion, oxidation, how to produce plutonium from spent fuels by describing various chemical treatment methods, which are currently used and were used in the past. It also contains characteristics of the purex reprocessing process which is the most widely used nowadays. And show processes to purify and metalize from recovered plutonium solution. Detection and analysis methods are introduced with key pints for handling, critical safety, toxicity, and effects on peoples. This report gives not only a general idea on what plutonium is, rather than deep technical description, but also basic knowledge on plutonium production and safeguards diversion from the view point of nonproliferation. 18 refs. (Author) .new.

  16. Chemistry research and development progress report, May-October, 1978

    International Nuclear Information System (INIS)

    Work in progress includes: calorimetry and thermodynamics of nuclear materials; americium recovery and purification; optimization of the cation exchange process for recovering americium and plutonium from molten salt extraction residues, photochemical separations of actinides; advanced ion exchange materials and techniques; secondary actinide recovery; removal of plutonium from lathe coolant oil; evaluation of tributyl phosphate-impregnated sorbent for plutonium-uranium separations; plutonium recovery in advance size reduction facility; plutonium peroxide precipitation; decontamination of Rocky Flats soil; soil decontamination at other Department of Energy sites; recovery of actinides from combustible wastes; induction-heated, tilt-pour furnace; vacuum melting; determination of plutonium and americium in salts and alloys by calorimetry; plutonium peroxide precipitation process; silica removal study; a comparative study of annular and Raschig ring-filled tanks; recovery of plutonium and americium from a salt cleanup alloy; and process development for recovery of americium from vacuum melt furnace crucibles

  17. Chemistry research and development progress report, May-October, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Miner, F. J.

    1979-08-30

    Work in progress includes: calorimetry and thermodynamics of nuclear materials; americium recovery and purification; optimization of the cation exchange process for recovering americium and plutonium from molten salt extraction residues, photochemical separations of actinides; advanced ion exchange materials and techniques; secondary actinide recovery; removal of plutonium from lathe coolant oil; evaluation of tributyl phosphate-impregnated sorbent for plutonium-uranium separations; plutonium recovery in advance size reduction facility; plutonium peroxide precipitation; decontamination of Rocky Flats soil; soil decontamination at other Department of Energy sites; recovery of actinides from combustible wastes; induction-heated, tilt-pour furnace; vacuum melting; determination of plutonium and americium in salts and alloys by calorimetry; plutonium peroxide precipitation process; silica removal study; a comparative study of annular and Raschig ring-filled tanks; recovery of plutonium and americium from a salt cleanup alloy; and process development for recovery of americium from vacuum melt furnace crucibles.

  18. Delocalization and new phase in Americium: theory

    Energy Technology Data Exchange (ETDEWEB)

    Soderlind, P

    1999-04-23

    Density-functional electronic structure calculations have been used to investigate the high pressure behavior of Am. At about 80 kbar (8 GPa) calculations reveal a monoclinic phase similar to the ground state structure of plutonium ({alpha}-Pu). The experimentally suggested {alpha}-U structure is found to be substantially higher in energy. The phase transition from fcc to the low symmetry structure is shown to originate from a drastic change in the nature of the electronic structure induced by the elevated pressure. A calculated volume collapse of about 25% is associated with the transition. For the low density phase, an orbital polarization correction to the local spin density (LSD) theory was applied. Gradient terms of the electron density were included in the calculation of the exchange/correlation energy and potential, according to the generalized gradient approximation (GGA). The results are consistent with a Mott transition; the 5f electrons are delocalized and bonding on the high density side of the transition and chemically inert and non-bonding (localized) on the other. Theory compares rather well with recent experimental data which implies that electron correlation effects are reasonably modeled in our orbital polarization scheme.

  19. PREPARATION OF PLUTONIUM HALIDES

    Science.gov (United States)

    Davidson, N.R.; Katz, J.J.

    1958-11-01

    A process ls presented for the preparation of plutonium trihalides. Plutonium oxide or a compound which may be readily converted to plutonlum oxide, for example, a plutonium hydroxide or plutonlum oxalate is contacted with a suitable halogenating agent. Speciflc agents mentioned are carbon tetrachloride, carbon tetrabromide, sulfur dioxide, and phosphorus pentachloride. The reaction is carried out under superatmospberic pressure at about 300 icient laborato C.

  20. PLUTONIUM-ZIRCONIUM ALLOYS

    Science.gov (United States)

    Schonfeld, F.W.; Waber, J.T.

    1960-08-30

    A series of nuclear reactor fuel alloys consisting of from about 5 to about 50 at.% zirconium (or higher zirconium alloys such as Zircaloy), balance plutonium, and having the structural composition of a plutonium are described. Zirconium is a satisfactory diluent because it alloys readily with plutonium and has desirable nuclear properties. Additional advantages are corrosion resistance, excellent fabrication propenties, an isotropie structure, and initial softness.