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Sample records for plutonium oxide powder

  1. Effects of pressure on thermal transport in plutonium oxide powder

    International Nuclear Information System (INIS)

    Bielenberg, Patricia; Prenger, F. Coyne; Veirs, Douglas Kirk; Jones, Jerry

    2004-01-01

    Radial temperature profiles in plutonium oxide (PuO 2 ) powder were measured in a cylindrical vessel over a pressure range of 0.055 to 334.4 kPa with two different fill gases, helium and argon. The fine PuO 2 powder provides a very uniform self-heating medium amenable to relatively simple mathematical descriptions. At low pressures ( 2 powder has small particle sizes (on the order of 1 to 10 μm), random particle shapes, and high porosity so a more general model was required for this system. The model correctly predicts the temperature profiles of the powder over the wide pressure range for both argon and helium as fill gases. The effective thermal conductivity of the powder bed exhibits a pressure dependence at higher pressures because the pore sizes in the interparticle contact area are relatively small (less than 1 μm) and the Knudsen number remains above the continuum limit at these conditions for both fill gases. Also, the effective thermal conductivity with argon as a fill gas is higher than expected at higher pressures because the solid pathways account for over 80% of the effective powder conductivity. The results obtained from this model help to bring insight to the thermal conductivity of very fine ceramic powders with different fill gases.

  2. Investigation on the attribute detection of plutonium oxide

    International Nuclear Information System (INIS)

    Liu Suping; Hao Fanhua; Gong Jian; Wu Huailong; Hu Guangchun; Hu Yongbo

    2006-01-01

    A long-term detection of the 871 keV gamma-ray emitted from two PuO 2 samples (100 g recovered powder and 8 g pure powder) has been conduced with a high resolution HPGe spectrometer in a lead-shielded chamber. An on-site detection of the 871 keV gamma-ray emitted from 200 g recovered PuO 2 powder has also been performed. Although the 871 keV peak can be observed in all 3 samples, other characteristic gamma-rays related to the existence of oxygen are not seen. The 871 keV peak is absent in spectra acquired from metal plutonium samples, which can answer whether the plutonium material in a sealed container is metal plutonium or plutonium oxide, i.e. the 871 keV gamma-ray can be an evidence for the absence of metal plutonium. The 871 keV peak is strong in the spectrum of 200 g recovered PuO 2 sample. A software of plutonium abundance and age analysis developed by ourselves can identify the 871 keV peak from a spectrum of 10 min. If the system is appropriately improved, it can be used to detect attribute of plutonium oxide through identifying 871 keV gamma-ray peak. (authors)

  3. Pressurization of Containment Vessels from Plutonium Oxide Contents

    International Nuclear Information System (INIS)

    Hensel, S.

    2012-01-01

    Transportation and storage of plutonium oxide is typically done using a convenience container to hold the oxide powder which is then placed inside a containment vessel. Intermediate containers which act as uncredited confinement barriers may also be used. The containment vessel is subject to an internal pressure due to several sources including; (1) plutonium oxide provides a heat source which raises the temperature of the gas space, (2) helium generation due to alpha decay of the plutonium, (3) hydrogen generation due to radiolysis of the water which has been adsorbed onto the plutonium oxide, and (4) degradation of plastic bags which may be used to bag out the convenience can from a glove box. The contributions of these sources are evaluated in a reasonably conservative manner.

  4. Comparison of physical chemical properties of powders and respirable aerosols of industrial mixed uranium and plutonium oxide fuels

    International Nuclear Information System (INIS)

    Eidson, A.F.

    1982-01-01

    Studies were performed to characterize physical and chemical properties which may be important in determining the metabolism of accidentally released, inhaled aerosols of industrial mixed uranium and plutonium oxide fuels and to compare the properties of bulk powders and the respirable fraction they include. X-ray diffraction measurements showed that analysis of mixed-oxide powders from four process steps served to characterize their respirable fractions. IR spectroscopy was useful as a method to detect organic binders that were not observed by X-ray diffraction methods. Both X-ray diffraction and IR spectroscopy methods can be used in combination to identify the sources of a complex aerosol that might be released from more than one fabrication step. Isotopic distributions in powders and aerosols showed that information important for radiation dose to tissue calculations or Pu lung burden estimates can be obtained by analysis of powders. (U.K.)

  5. Materials identification and surveillance project item evaluation: Items, impure plutonium oxide (ATL27960) and pure plutonium oxide (PEOR3258)

    International Nuclear Information System (INIS)

    Allen, T.; Appert, Q.; Davis, C.

    1997-03-01

    In this report, Los Alamos scientists characterize properties relevant to storage of an impure plutonium oxide (74 mass % plutonium) in accordance with the Department of Energy (DOE) standard DOE-STD-3013-96. This oxide is of interest because it is the first impure plutonium oxide sample to be evaluated and it is similar to other materials that must be stored. Methods used to characterize the oxide at certain points during calcination include surface-area analyses, mass loss-on-ignition (LOI) measurements, elemental analysis, moisture-adsorption measurements, and quantitative supercritical-CO 2 extraction of adsorbed water. Significant decreases in the LOI and surface area occurred as the oxide was calcined at progressively increasing temperatures. Studies indicate that supercritical-CO 2 extraction is an effective method for removing adsorbed water from oxides. We extracted the water from powdered oxides (high-purity ZrO 2 , pure PuO 2 , and impure plutonium oxide) using CO 2 at 3000 psi pressure and 75 degrees C, and we quantitatively determined it by using gravimetric and dew-point procedures. The effectiveness of the extraction method is demonstrated by good agreement between the amounts of water extracted from pure zirconium and plutonium dioxides and the mass changes obtained from LOI analyses. However, the amount of moisture (0.025 mass %) extracted from the impure plutonium oxide after it had been calcined at 950 degrees C and stored for a period of months is much less than the LOI value (0.97 mass %). These results imply that the impure plutonium oxide is free of adsorbed water after calcination at 950 degrees C, even though the sample does not satisfy the LOI requirement of <0.50 mass % for storage

  6. Thermal and Physical Properties of Plutonium Dioxide Produced from the Oxidation of Metal: a Data Summary

    Energy Technology Data Exchange (ETDEWEB)

    Wayne, David M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-01-13

    The ARIES Program at the Los Alamos National Laboratory removes plutonium metal from decommissioned nuclear weapons, and converts it to plutonium dioxide in a specially-designed Direct Metal Oxidation furnace. The plutonium dioxide is analyzed for specific surface area, particle size distribution, and moisture content. The purpose of these analyses is to certify that the plutonium dioxide powder meets or exceeds the specifications of the end-user, and the specifications for the packaging and transport of nuclear materials. Analytical results from plutonium dioxide from ARIES development activities, from ARIES production activities, from muffle furnace oxidation of metal, and from metal that was oxidized over a lengthy time interval in air at room temperature, are presented. The processes studied produce plutonium dioxide powder with distinct differences in measured properties, indicating the significant influence of oxidation conditions on physical properties.

  7. Precipitation of plutonium (III) oxalate and calcination to plutonium oxide

    International Nuclear Information System (INIS)

    Esteban, A.; Orosco, E.H.; Cassaniti, P.; Greco, L.; Adelfang, P.

    1989-01-01

    The plutonium based fuel fabrication requires the conversion of the plutonium nitrate solution from nuclear fuel reprocessing into pure PuO2. The conversion method based on the precipitation of plutonium (III) oxalate and subsequent calcination has been studied in detail. In this procedure, plutonium (III) oxalate is precipitated, at room temperature, by the slow addition of 1M oxalic acid to the feed solution, containing from 5-100 g/l of plutonium in 1M nitric acid. Before precipitation, the plutonium is adjusted to trivalent state by addition of 1M ascorbic acid in the presence of an oxidation inhibitor such as hydrazine. Finally, the precipitate is calcinated at 700 deg C to obtain PuO2. A flowsheet is proposed in this paper including: a) A study about the conditions to adjust the plutonium valence. b) Solubility data of plutonium (III) oxalate and measurements of plutonium losses to the filtrate and wash solution. c) Characterization of the obtained products. Plutonium (III) oxalate has several potential advantages over similar conversion processes. These include: 1) Formation of small particle sizes powder with good pellets fabrication characteristics. 2) The process is rather insensitive to most process variables, except nitric acid concentration. 3) Ambient temperature operations. 4) The losses of plutonium to the filtrate are less than in other conversion processes. (Author) [es

  8. Gas adsorption during storage of plutonium dioxide powders

    International Nuclear Information System (INIS)

    Cuillerdier, C.; Cossonnet, C.; Germain, M.

    1984-10-01

    Adsorption phenomena occuring in plutonium dioxide containers are studied for the determination of safe conditions for storage and transportation of plutonium dioxide powders. Adsorption on dried PuO 2 of air individual gases, influence of powder isotopic composition, chemisorption, effect of moisture are determined. Adsorption of dry air obeys an Elovich's law for its kinetics it is greatly exchange by α radiolysis. Pressure in the container can be reduced by storage under dry inert gas (Ar), decreasing the PuO 2 load and using powder containing preadsorbed water or wet air then radiolysis may occur (H 2 formation)

  9. Synthesis of complex plutonium oxides with alkaline-earth metals

    International Nuclear Information System (INIS)

    Suzuki, Yasufumi; Nakajima, Kunihisa; Iwai, Takashi; Ohmichi, Toshihiko; Yamawaki, Michio.

    1995-03-01

    Complex plutonium(IV) oxides with strontium and barium, SuPuO 3 and BaPuO 3 , were synthesized and their crystal structure was analyzed. Compacted mixture of plutonium dioxide powder and the carbonate of strontium or barium was heated in a stream of argon gas using a cell with a small orifice. The products obtained were found to be composed of a nearly single phase showing the structure of orthorhombic slightly distorted from cubic. (author)

  10. Method to manufacture a nuclear fuel from uranium-plutonium monocarbide or uranium-plutonium mononitride

    International Nuclear Information System (INIS)

    Krauth, A.; Mueller, N.

    1977-01-01

    Pure uranium carbide or nitride is converted with plutonium oxide and carbon (all in powder form) to uranium-plutonium monocarbide or mononitride by cold pressing and sintering at about 1600 0 C. Pure uranium carbide or uranium nitride powder is firstly prepared without extensive safety measures. The pure uranium carbide or nitride powder can also be inactivated by using chemical substances (e.g. stearic acid) and be handled in air. The sinterable uranium carbide or nitride powder (or also granulate) is then introduced into the plutonium line and mixed with a nonstoichiometrically adjusted, prereacted mixture of plutonium oxide and carbon, pressed to pellets and reaction sintered. The surface of the uranium-plutonium carbide (higher metal content) can be nitrated towards the end of the sinter process in a stream of nitrogen. The protective layer stabilizes the carbide against the water and oxygen content in air. (IHOE) [de

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

    International Nuclear Information System (INIS)

    Eidson, A.F.; Mewhinney, J.A.

    1978-01-01

    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 UO x and PuO x 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)

  12. Plutonium oxides and uranium and plutonium mixed oxides. Carbon determination

    International Nuclear Information System (INIS)

    Anon.

    Determination of carbon in plutonium oxides and uranium plutonium mixed oxides, suitable for a carbon content between 20 to 3000 ppm. The sample is roasted in oxygen at 1200 0 C, the carbon dioxide produced by combustion is neutralized by barium hydroxide generated automatically by coulometry [fr

  13. Studies on O/M ratio determination in uranium oxide, plutonium oxide and uranium-plutonium mixed oxide

    International Nuclear Information System (INIS)

    Sampath, S.; Chawla, K.L.

    1975-01-01

    Thermogravimetric studies were carried out in unsintered and sintered samples of uranium oxide, plutonium oxide and uranium-plutonium mixed oxide under different atmospheric conditions (air, argon and moist argon/hydrogen). Moisture loss was found to occur below 200 0 C for uranium dioxide samples, upto 700 0 C for sintered plutonium dioxide and negligible for sintered samples. The O/M ratios for non-stoichiometric uranium dioxide (sintered and unsintered), plutonium dioxide and mixed uranium and plutonium oxides (sintered) could be obtained with a precision of +- 0.002. Two reference states UOsub(2.000) and UOsub(2.656) were obtained for uranium dioxide and the reference state MOsub(2.000) was used for other cases. For unsintered plutonium dioxide samples, accurate O/M ratios could not be obtained of overlap of moisture loss with oxygen loss/gain. (author)

  14. Plutonium Immobilization Project Concept for Dustless Transfer of Powder

    International Nuclear Information System (INIS)

    Ward, C.R.

    2001-01-01

    Plutonium powder will be brought into the Plutonium Immobilization Plant in Food Pack Cans in 3013 packages. The Food Pack Cans will be removed from the 3013 outer and inner can. This document describes their concept and completes PIP milestone 2.2.3.4/FY01/c, Complete Concept for Material Transfer

  15. Fabrication of mixed oxide fuel using plutonium from dismantled weapons

    International Nuclear Information System (INIS)

    Blair, H.T.; Chidester, K.; Ramsey, K.B.

    1996-01-01

    A very brief summary is presented of experimental studies performed to support the use of plutonium from dismantled weapons in fabricating mixed oxide (MOX) fuel for commercial power reactors. Thermal treatment tests were performed on plutonium dioxide powder to determine if an effective dry gallium removal process could be devised. Fabrication tests were performed to determine the effects of various processing parameters on pellet quality. Thermal tests results showed that the final gallium content is highly dependent on the treatment temperature. Fabrication tests showed that the milling process, sintering parameters, and uranium feed did effect pellet properties. 1 ref., 1 tab

  16. Ultra-small plutonium oxide nanocrystals: an innovative material in plutonium science.

    Science.gov (United States)

    Hudry, Damien; Apostolidis, Christos; Walter, Olaf; Janssen, Arne; Manara, Dario; Griveau, Jean-Christophe; Colineau, Eric; Vitova, Tonya; Prüssmann, Tim; Wang, Di; Kübel, Christian; Meyer, Daniel

    2014-08-11

    Apart from its technological importance, plutonium (Pu) is also one of the most intriguing elements because of its non-conventional physical properties and fascinating chemistry. Those fundamental aspects are particularly interesting when dealing with the challenging study of plutonium-based nanomaterials. Here we show that ultra-small (3.2±0.9 nm) and highly crystalline plutonium oxide (PuO2 ) nanocrystals (NCs) can be synthesized by the thermal decomposition of plutonyl nitrate ([PuO2 (NO3 )2 ]⋅3 H2 O) in a highly coordinating organic medium. This is the first example reporting on the preparation of significant quantities (several tens of milligrams) of PuO2 NCs, in a controllable and reproducible manner. The structure and magnetic properties of PuO2 NCs have been characterized by a wide variety of techniques (powder X-ray diffraction (PXRD), X-ray absorption fine structure (XAFS), X-ray absorption near edge structure (XANES), TEM, IR, Raman, UV/Vis spectroscopies, and superconducting quantum interference device (SQUID) magnetometry). The current PuO2 NCs constitute an innovative material for the study of challenging problems as diverse as the transport behavior of plutonium in the environment or size and shape effects on the physics of transuranium elements. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  17. Process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide

    International Nuclear Information System (INIS)

    Heremanns, R.H.; Vandersteene, J.J.

    1983-01-01

    The invention concerns a process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide in the form of PuO 2 . Mixed fuels consisting of uranium oxide and plutonium oxide are being used more and more. The plants which prepare these mixed fuels have around 5% of the total mass of fuels as fabrication residue, either as waste or scrap. In view of the high cost of plutonium, it has been attempted to recover this plutonium from the fabrication residues by a process having a purchase price lower than the price of plutonium. The problem is essentially to separate the plutonium, the uranium and the impurities. The residues are fluorinated, the UF 6 and PuF 6 obtained are separated by selective absorption of the PuF 6 on NaF at a temperature of at least 400 0 C, the complex obtained by this absorption is dissolved in nitric acid solution, the plutonium is precipitated in the form of plutonium oxalate by adding oxalic acid, and the precipitated plutonium oxalate is calcined

  18. Criticality safety analysis for plutonium dissolver using silver mediated electrolytic oxidation method

    International Nuclear Information System (INIS)

    Umeda, Miki; Sugikawa, Susumu; Nakamura, Kazuhito; Egashira, Tetsurou

    1998-08-01

    Design and construction of a plutonium dissolver using silver mediated electrolytic oxidation method are promoted in NUCEF. Criticality safety analysis for the plutonium dissolver is described in this report. The electrolytic plutonium dissolver consists of connection pipes and three pots for MOX powder supply, circulation and electrolysis. The criticality control for the dissolver is made by geometrically safe shape with mass limitation. Monte Carlo code KENO-IV using MGCL-137 library based on ENDF/B-IV was used for the criticality safety analysis for the plutonium dissolver. Considering the required size for construction and criticality safety, diameter of pot and distance between two pots were determined. On this condition, the criticality safety analysis for the plutonium dissolver with connection pipes was carried out. As the result of the criticality safety analysis, an effective neutron multiplication factor keff of 0.91 was obtained and the criticality safety of the plutonium dissolver was confirmed on the basis of criteria of ≤0.95. (author)

  19. Validation of KENO, ANISN and Hansen-Roach cross-section set on plutonium oxide and metal fuel system

    International Nuclear Information System (INIS)

    Matsumoto, Tadakuni; Yumoto, Ryozo; Nakano, Koh.

    1980-01-01

    In the previous report, the authors discussed the validity of KENO, ANISN and Hansen-Roach 16 group cross-section set on the critical plutonium nitrate solution systems with various geometries, absorbers and neutron interactions. The purpose of the present report is to examine the validity of the same calculation systems on the homogeneous plutonium oxide and plutonium-uranium mixed oxide fuels with various density values. Eleven experiments adopted for validation are summarized. First six experiments were performed at Pacific Northwest Laboratory of Battelle Memorial Institute, and the remaining five at Los Alamos Scientific Laboratory. The characteristics of core fuel are given, and the isotopic composition of plutonium, the relation between H/(Pu + U) atomic ratio and fuel density as compared with the atomic ratios of PuO 2 and mixed oxides in powder storage and pellet fabrication processes, and critical core dimensions and reflector conditions are shown. The effective multiplication factors were calculated with the KENO code. In case of the metal fuels with simple sphere geometry, additional calculations with the ANISN code were performed. The criticality calculation system composed of KENO, ANISN and Hansen-Roach cross-section set was found to be valid for calculating the criticality on plutonium oxide, plutonium-uranium mixed oxide, plutonium metal and uranium metal fuel systems as well as on plutonium solution systems with various geometries, absorbers and neutron interactions. There seems to remain some problems in the method for evaluating experimental correction. Some discussions foloow. (Wakatsuki, Y.)

  20. Plutonium Oxide Process Capability Work Plan

    Energy Technology Data Exchange (ETDEWEB)

    Meier, David E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tingey, Joel M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-02-28

    Pacific Northwest National Laboratory (PNNL) has been tasked to develop a Pilot-scale Plutonium-oxide Processing Unit (P3U) providing a flexible capability to produce 200g (Pu basis) samples of plutonium oxide using different chemical processes for use in identifying and validating nuclear forensics signatures associated with plutonium production. Materials produced can also be used as exercise and reference materials.

  1. Chemical and Radiochemical Composition of Thermally Stabilized Plutonium Oxide from the Plutonium Finishing Plant Considered as Alternate Feedstock for the Mixed Oxide Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Tingey, Joel M.; Jones, Susan A.

    2005-01-01

    Eighteen plutonium oxide samples originating from the Plutonium Finishing Plant (PFP) on the Hanford Site were analyzed to provide additional data on the suitability of PFP thermally stabilized plutonium oxides and Rocky Flats oxides as alternate feedstock to the Mixed Oxide Fuel Fabrication Facility (MFFF). Radiochemical and chemical analyses were performed on fusions, acid leaches, and water leaches of these 18 samples. The results from these destructive analyses were compared with nondestructive analyses (NDA) performed at PFP and the acceptance criteria for the alternate feedstock. The plutonium oxide materials considered as alternate feedstock at Hanford originated from several different sources including Rocky Flats oxide, scrap from the Remote Mechanical C-Line (RMC) and the Plutonium Reclamation Facility (PRF), and materials from other plutonium conversion processes at Hanford. These materials were received at PFP as metals, oxides, and solutions. All of the material considered as alternate feedstock was converted to PuO2 and thermally stabilized by heating the PuO2 powder at 950 C in an oxidizing environment. The two samples from solutions were converted to PuO2 by precipitation with Mg(OH)2. The 18 plutonium oxide samples were grouped into four categories based on their origin. The Rocky Flats oxide was divided into two categories, low- and high-chloride Rocky Flats oxides. The other two categories were PRF/RMC scrap oxides, which included scrap from both process lines and oxides produced from solutions. The two solution samples came from samples that were being tested at Pacific Northwest National Laboratory because all of the plutonium oxide from solutions at PFP had already been processed and placed in 3013 containers. These samples originated at the PFP and are from plutonium nitrate product and double-pass filtrate solutions after they had been thermally stabilized. The other 16 samples originated from thermal stabilization batches before canning at

  2. Tabulated Neutron Emission Rates for Plutonium Oxide

    Energy Technology Data Exchange (ETDEWEB)

    Shores, Erik Frederick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-07-24

    This work tabulates neutron emission rates for 80 plutonium oxide samples as reported in the literature. Plutonium-­238 and plutonium-­239 oxides are included and such emission rates are useful for scaling tallies from Monte Carlo simulations and estimating dose rates for health physics applications.

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

    International Nuclear Information System (INIS)

    1998-06-01

    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

  4. Plutonium oxide dissolution

    International Nuclear Information System (INIS)

    Gray, J.H.

    1992-01-01

    Several processing options for dissolving plutonium oxide (PuO 2 ) from high-fired materials have been studied. The scoping studies performed on these options were focused on PuO 2 typically generated by burning plutonium metal and PuO 2 produced during incineration of alpha contaminated waste. At least two processing options remain applicable for dissolving high-fired PuO 2 in canyon dissolvers. The options involve solid solution formation of PuO 2 With uranium oxide (UO 2 ) and alloying incinerator ash with aluminum. An oxidative dissolution process involving nitric acid solutions containing a strong oxidizing agent, such as cerium (IV), was neither proven nor rejected. This uncertainty was due to difficulty in regenerating cerium (IV) ions during dissolution. However, recent work on silver-catalyzed dissolution of PuO 2 with persulfate has demonstrated that persulfate ions regenerate silver (II). Use of persulfate to regenerate cerium (IV) or bismuth (V) ions during dissolution of PuO 2 materials may warrant further study

  5. Analysis of pressurization of plutonium oxide storage vials during a postulated fire

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, J.; Kesterson, M.; Hensel, S.

    2015-02-10

    The documented safety analysis for the Savannah River Site evaluates the consequences of a postulated 1000 °C fire in a glovebox. The radiological dose consequences for a pressurized release of plutonium oxide powder during such a fire depend on the maximum pressure that is attained inside the oxide storage vial. To enable evaluation of the dose consequences, pressure transients and venting flow rates have been calculated for exposure of the storage vial to the fire. A standard B vial with a capacity of approximately 8 cc was selected for analysis. The analysis compares the pressurization rate from heating and evaporation of moisture adsorbed onto the plutonium oxide contents of the vial with the pressure loss due to venting of gas through the threaded connection between the vial cap and body. Tabulated results from the analysis include maximum pressures, maximum venting velocities, and cumulative vial volumes vented during the first 10 minutes of the fire transient. Results are obtained for various amounts of oxide in the vial, various amounts of adsorbed moisture, different vial orientations, and different surface fire exposures.

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

    International Nuclear Information System (INIS)

    Jarbo, G.J.; Faught, P.; Hildebrandt, B.

    1980-05-01

    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)

  7. International shipment of plutonium by air

    International Nuclear Information System (INIS)

    Mercado, J.E.; McGrogan, J.P.

    1995-05-01

    In support of the United States (US) Government's decision to place excess plutonium oxide at the US Department of Energy's (DOE) Hanford Site under International Atomic Energy Agency (IAEA) safeguards, the Department of State notified the Congress that a plutonium storage vault at the Plutonium Finishing Plant at the Hanford Site would be added to the eligible facilities list. As part of the preparations to transfer the plutonium oxide under IAEA safeguards, samples of the powder were taken from the inventory to be shipped to the IAEA headquarters in Vienna, Austria, for laboratory analysis. The analysis of these samples was of high priority, and the IAEA requested that the material be shipped by aircraft, the most expeditious method

  8. HB-Line Plutonium Oxide Data Collection Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Watkins, R. [Savannah River Nuclear Solutions; Varble, J. [Savannah River Nuclear Solutions; Jordan, J. [Savannah River Nuclear Solutions

    2015-05-26

    HB-Line and H-Canyon will handle and process plutonium material to produce plutonium oxide for feed to the Mixed Oxide Fuel Fabrication Facility (MFFF). However, the plutonium oxide product will not be transferred to the MFFF directly from HB-Line until it is packaged into a qualified DOE-STD-3013-2012 container. In the interim, HB-Line will load plutonium oxide into an inner, filtered can. The inner can will be placed in a filtered bag, which will be loaded into a filtered outer can. The outer can will be loaded into a certified 9975 with getter assembly in compliance with onsite transportation requirement, for subsequent storage and transfer to the K-Area Complex (KAC). After DOE-STD-3013-2012 container packaging capabilities are established, the product will be returned to HB-Line to be packaged into a qualified DOE-STD-3013-2012 container. To support the transfer of plutonium oxide to KAC and then eventually to MFFF, various material and packaging data will have to be collected and retained. In addition, data from initial HB-Line processing operations will be needed to support future DOE-STD-3013-2012 qualification as amended by the HB-Line DOE Standard equivalency. As production increases, the volume of data to collect will increase. The HB-Line data collected will be in the form of paper copies and electronic media. Paper copy data will, at a minimum, consist of facility procedures, nonconformance reports (NCRs), and DCS print outs. Electronic data will be in the form of Adobe portable document formats (PDFs). Collecting all the required data for each plutonium oxide can will be no small effort for HB-Line, and will become more challenging once the maximum annual oxide production throughput is achieved due to the sheer volume of data to be collected. The majority of the data collected will be in the form of facility procedures, DCS print outs, and laboratory results. To facilitate complete collection of this data, a traveler form will be developed which

  9. Oxidation-state maxima in plutonium chemistry

    International Nuclear Information System (INIS)

    Silver, G.L.

    2013-01-01

    Maxima in the fractions of the trivalent and hexavalent oxidation states of plutonium are inherent in the algebra of its disproportionation reactions. The maxima do not support overall disproportionation equations as satisfactory representations of aqueous plutonium. (author)

  10. A review of plutonium oxalate decomposition reactions and effects of decomposition temperature on the surface area of the plutonium dioxide product

    International Nuclear Information System (INIS)

    Orr, R.M.; Sims, H.E.; Taylor, R.J.

    2015-01-01

    Plutonium (IV) and (III) ions in nitric acid solution readily form insoluble precipitates with oxalic acid. The plutonium oxalates are then easily thermally decomposed to form plutonium dioxide powder. This simple process forms the basis of current industrial conversion or ‘finishing’ processes that are used in commercial scale reprocessing plants. It is also widely used in analytical or laboratory scale operations and for waste residues treatment. However, the mechanisms of the thermal decompositions in both air and inert atmospheres have been the subject of various studies over several decades. The nature of intermediate phases is of fundamental interest whilst understanding the evolution of gases at different temperatures is relevant to process control. The thermal decomposition is also used to control a number of powder properties of the PuO_2 product that are important to either long term storage or mixed oxide fuel manufacturing. These properties are the surface area, residual carbon impurities and adsorbed volatile species whereas the morphology and particle size distribution are functions of the precipitation process. Available data and experience regarding the thermal and radiation-induced decompositions of plutonium oxalate to oxide are reviewed. The mechanisms of the thermal decompositions are considered with a particular focus on the likely redox chemistry involved. Also, whilst it is well known that the surface area is dependent on calcination temperature, there is a wide variation in the published data and so new correlations have been derived. Better understanding of plutonium (III) and (IV) oxalate decompositions will assist the development of more proliferation resistant actinide co-conversion processes that are needed for advanced reprocessing in future closed nuclear fuel cycles. - Highlights: • Critical review of plutonium oxalate decomposition reactions. • New analysis of relationship between SSA and calcination temperature. • New SEM

  11. Calibration of UFBC counters and their performance in the assay of large mass plutonium samples

    International Nuclear Information System (INIS)

    Verrecchia, G.P.D.; Smith, B.G.R.; Cranston, R.

    1991-01-01

    This paper reports on the cross-calibration of four Universal Fast Breeder reactor assembly coincidence (UFBC) counters using multi-can containers of Plutonium oxide powders with masses between 2 and 12 Kg of plutonium and a parametric study on the sensitivity of the detector response to the positioning or removal and substitution of the material with empty cans. The paper also reports on the performance of the UFBC for routine measurements on large mass, multi-can containers of plutonium oxide powders and compares the results to experience previously obtained in the measurement of fast reactor type fuel assemblies in the mass range 2 to 16 Kg of plutonium

  12. CSER 01-008 Canning of Thermally Stabilized Plutonium Oxide Powder in PFP Glovebox HC-21A

    International Nuclear Information System (INIS)

    ERICKSON, D.G.

    2001-01-01

    This document presents the analysis performed to support the canning operation in HC-21A. Most of the actual analysis was performed for the operation in HC-18M and HA-20MB, and is documented in HNF-2707 Rev I a (Erickson 2001a). This document will reference Erickson (2001a) as necessary to support the operation in HC-21A. The plutonium stabilization program at the Plutonium Finishing Plant (PFP) uses heat to convert plutonium-bearing materials into dry powder that is chemically stable for long term storage. The stabilized plutonium is transferred into one of several gloveboxes for the canning process, Gloveboxes HC-18M in Room 228'2, HA-20MB in Room 235B, and HC-21A in Room 230B are to be used for this process. This document presents the analysis performed to support the canning operation in HC-21A. Most of the actual analysis was performed for the operation in HC-I8M and HA-20MB, and is documented in HNF-2707 Rev l a (Erickson 2001a). This document will reference Erickson (2001a) as necessary to support the operation in HC-21A. Evaluation of this operation included normal, base cases, and contingencies. The base cases took the normal operations for each type of feed material and added the likely off-normal events. Each contingency is evaluated assuming the unlikely event happens to the conservative base case. Each contingency was shown to meet the double contingency requirement. That is, at least two unlikely, independent, and concurrent changes in process conditions are required before a criticality is possible

  13. Precipitation of plutonium from acidic solutions using magnesium oxide

    International Nuclear Information System (INIS)

    Jones, S.A.

    1994-01-01

    Magnesium oxide will be used as a neutralizing agent for acidic plutonium-containing solutions. It is expected that as the magnesium oxide dissolves, the pH of the solution will rise, and plutonium will precipitate. The resulting solid will be tested for suitability to storage. The liquid is expected to contain plutonium levels that meet disposal limit requirements

  14. Solubility of plutonium dioxide aerosols, in vitro

    International Nuclear Information System (INIS)

    Newton, G.J.; Kanapilly, G.M.

    1976-01-01

    Solubility of plutonium aerosols is an important parameter in establishing risk estimates for industrial workers who might accidentally inhale these materials and in evaluating environmental health impacts associated with Pu. In vitro solubility of industrial plutonium aerosols in a simulated lung fluid is compared to similar studies with ultrafine aerosols from laser ignition of delta phase plutonium metal and laboratory-produced spherical particles of 238 PuO 2 and 239 PuO 2 . Although relatively insoluble, industrial plutonium-mixed oxide aerosols were much more soluble than laboratory-produced plutonium dioxide particles. Chain agglomerate aerosols from laser ignition of metallic Pu indicated in vitro dissolution half-times of 10 and 50 days for activity median aerodynamic diameter (AMAD) of 0.7 and 2.3 μm, respectively. Plutonium-containing mixed oxide aerosols indicated dissolution half-times of 40 to 500 days for particles formed by industrial powder comminution and blending. Centerless grinding of fuel pellets yielded plutonium-containing aerosols with dissolution half-times of 1200 to 8000 days. All mixed oxide particles were in the size range 1.0 μm to 2.5 μm AMAD

  15. Radiochemical procedure for the determination of plutonium isotopes in powdered milk

    International Nuclear Information System (INIS)

    Taddei, M.H.T.; Silva, N.C.

    2006-01-01

    A radiochemical procedure for the determination of alpha-emitting isotopes of plutonium in powdered milk is proposed. The procedure involves sample dissolution (by HNO 3 and HClO 4 ), separation by ionic-exchange resin, electrodeposition and alpha-spectroscopy. In order to determine the chemical recovery, 242 Pu was employed as a tracer. A reference material (Marine Sediment IAEA 135) was analyzed to validate such procedure, and to show its reliability. Afterwards, some powdered milk, produced for international trade, was analyzed and chemical recovery was found to be around 95%. (author)

  16. Ni(III) - a new oxidant for the potentiometric determination of plutonium

    International Nuclear Information System (INIS)

    Gopinath, N.; Rama Rao, G.A.; Natarajan, P.R.

    1986-01-01

    Ni(III) was employed as an oxidant for the quantitative oxidation of plutonium to its hexavalent state and this makes possible to determine plutonium concentrations potentiometrically by the Fe(II)-dichromate redox titration method. The unreacted oxidant is completely reduced by sulphamic acid prior to the determination of plutonium. (author)

  17. A review of plutonium oxalate decomposition reactions and effects of decomposition temperature on the surface area of the plutonium dioxide product

    Energy Technology Data Exchange (ETDEWEB)

    Orr, R.M.; Sims, H.E.; Taylor, R.J., E-mail: robin.j.taylor@nnl.co.uk

    2015-10-15

    Plutonium (IV) and (III) ions in nitric acid solution readily form insoluble precipitates with oxalic acid. The plutonium oxalates are then easily thermally decomposed to form plutonium dioxide powder. This simple process forms the basis of current industrial conversion or ‘finishing’ processes that are used in commercial scale reprocessing plants. It is also widely used in analytical or laboratory scale operations and for waste residues treatment. However, the mechanisms of the thermal decompositions in both air and inert atmospheres have been the subject of various studies over several decades. The nature of intermediate phases is of fundamental interest whilst understanding the evolution of gases at different temperatures is relevant to process control. The thermal decomposition is also used to control a number of powder properties of the PuO{sub 2} product that are important to either long term storage or mixed oxide fuel manufacturing. These properties are the surface area, residual carbon impurities and adsorbed volatile species whereas the morphology and particle size distribution are functions of the precipitation process. Available data and experience regarding the thermal and radiation-induced decompositions of plutonium oxalate to oxide are reviewed. The mechanisms of the thermal decompositions are considered with a particular focus on the likely redox chemistry involved. Also, whilst it is well known that the surface area is dependent on calcination temperature, there is a wide variation in the published data and so new correlations have been derived. Better understanding of plutonium (III) and (IV) oxalate decompositions will assist the development of more proliferation resistant actinide co-conversion processes that are needed for advanced reprocessing in future closed nuclear fuel cycles. - Highlights: • Critical review of plutonium oxalate decomposition reactions. • New analysis of relationship between SSA and calcination temperature.

  18. The disposition of civil plutonium in the UK

    International Nuclear Information System (INIS)

    Sadnicki, M.J.; Barker, F.

    2001-01-01

    This paper quantifies the likely future stockpile of UK separated plutonium, and reviews current UK policy. The current strategy of storing plutonium oxide powder is shown to be inconsistent with passivity and disposability objectives. Analysis also shows that there is little potential for use on a commercial basis of Mixed-Oxide (MOX) fuel to reduce the stockpile. Four plutonium immobilisation options are defined, with particular reference to non-proliferation goals. The resource costs of implementing these options are quantified, together with the resource costs of a programme of Government-subsidized MOX use. Immobilisation may offer a more cost-effective solution than a MOX fuel route. (author)

  19. Standard test method for plutonium assay by plutonium (III) diode array spectrophotometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This test method describes the determination of total plutonium as plutonium(III) in nitrate and chloride solutions. The technique is applicable to solutions of plutonium dioxide powders and pellets (Test Methods C 697), nuclear grade mixed oxides (Test Methods C 698), plutonium metal (Test Methods C 758), and plutonium nitrate solutions (Test Methods C 759). Solid samples are dissolved using the appropriate dissolution techniques described in Practice C 1168. The use of this technique for other plutonium-bearing materials has been reported (1-5), but final determination of applicability must be made by the user. The applicable concentration range for plutonium sample solutions is 10–200 g Pu/L. 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 appropria...

  20. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

    1997-01-01

    A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium. (author)

  1. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

    1997-06-01

    A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium

  2. Recommended plutonium release fractions from postulated fires. Final report

    International Nuclear Information System (INIS)

    Kogan, V.; Schumacher, P.M.

    1993-12-01

    This report was written at the request of EG ampersand G Rocky Flats, Inc. in support of joint emergency planning for the Rocky Flats Plant (RFP) by EG ampersand G and the State of Colorado. The intent of the report is to provide the State of Colorado with an independent assessment of any respirable plutonium releases that might occur in the event of a severe fire at the plant. Fire releases of plutonium are of interest because they have been used by EG ampersand G to determine the RFP emergency planning zones. These zones are based on the maximum credible accident (MCA) described in the RFP Final Environmental Impact Statement (FEIS) of 1980, that MCA is assumed to be a large airplane crashing into a RFP plutonium building.The objective of this report was first, to perform a worldwide literature review of relevant release experiments from 1960 to the present and to summarize those findings, and second, to provide recommendations for application of the experimental data to fire release analyses at Rocky Flats. The latter step requires translation between experimental and expected RFP accident parameters, or ''scaling.'' The parameters of particular concern are: quantities of material, environmental parameters such as the intensity of a fire, and the physico-chemical forms of the plutonium. The latter include plutonium metal, bulk plutonium oxide powder, combustible and noncombustible wastes contaminated with plutonium oxide powder, and residues from plutonium extraction processes

  3. Plutonium-containing aerosols found within containment enclosures in industrial mixed-oxide reactor fuel fabrication

    International Nuclear Information System (INIS)

    Newton, G.J.; Yeh, H.C.; Stanley, J.A.

    1977-01-01

    Mixed oxide (PuO 2 and UO 2 ) nuclear reactor fuel pellets are fabricated within safety enclosures at Babcock and Wilcox's Park Township site near Apollo, PA. Forty-two sample runs of plutonium-containing aerosols were taken from within glove boxes during routine industrial operations. A small, seven-stage cascade impactor and the Lovelace Aerosol Particle Separator (LAPS) were used to determine aerodynamic size distribution and gross alpha aerosol concentration. Powder comminution and blending produced aerosols with lognormal size distributions characterized by activity median aerodynamic diameters (AMAD) of 1.89 +- 0.33 μm, sigma/sub g/ = 1.62 +- 0.09 and a gross alpha aerosol concentration range of 0.1 to 150 nCi/l. Slug pressing and grinding produced aerosols of AMAD = 3.08 +- 0.1 μm, sigma/sub g/ = 1.53 +- 0.01 and AMAD = 2.26 +- 0.16 μm, sigma/sub g/ = 1.68 +- 0.20, respectively. Gross alpha aerosol concentrations ranged from 3.4 to 450 nCi/l. Centerless grinding produced similar-sized aerosols but the gross alpha concentration ranged from 220 to 1690 nCi/l. In vitro solubility studies on selected LAPS samples in a lung fluid simulant indicate that plutonium mixed-oxide aerosols are more soluble than laboratory-produced plutonium aerosols

  4. Ceramification: A plutonium immobilization process

    Energy Technology Data Exchange (ETDEWEB)

    Rask, W.C. [Dept. of Energy, Golden, CO (United States); Phillips, A.G. [Rocky Flats Environmental Technology Site, Golden, CO (United States)

    1996-05-01

    This paper describes a low temperature technique for stabilizing and immobilizing actinide compounds using a combination process/storage vessel of stainless steel, in which measured amounts of actinide nitrate solutions and actinide oxides (and/or residues) are systematically treated to yield a solid article. The chemical ceramic process is based on a coating technology that produces rare earth oxide coatings for defense applications involving plutonium. The final product of this application is a solid, coherent actinide oxide with process-generated encapsulation that has long-term environmental stability. Actinide compounds can be stabilized as pure materials for ease of re-use or as intimate mixtures with additives such as rare earth oxides to increase their degree of proliferation resistance. Starting materials for the process can include nitrate solutions, powders, aggregates, sludges, incinerator ashes, and others. Agents such as cerium oxide or zirconium oxide may be added as powders or precursors to enhance the properties of the resulting solid product. Additives may be included to produce a final product suitable for use in nuclear fuel pellet production. The process is simple and reduces the time and expense for stabilizing plutonium compounds. It requires a very low equipment expenditure and can be readily implemented into existing gloveboxes. The process is easily conducted with less associated risk than proposed alternative technologies.

  5. Pyro-oxidation of plutonium spent salts with sodium carbonate

    International Nuclear Information System (INIS)

    Bourges, G.; Godot, A.; Valot, C.; Devillard, D.

    2001-01-01

    The purification of plutonium generates spent salts, which are temporarily stored in a nuclear building. A development programme for pyrochemical treatment is in progress to stabilize and concentrate these salts in order to reduce the quantities for long-term disposal. The treatment, inspired by work previously done by LANL, consists of a pyro-oxidation of the salt with sodium carbonate to convert the actinides into oxides, then of a vacuum distillation to separate the oxides from the volatile salt matrix. Pyro-oxidation of NaCl/KCl base spent salts first produces a 'black salt' which contains more than 97% of the initial actinides. XRD analyses indicate PuO 2 as major plutonium species and sodium plutonates or plutonium sub-oxides PuO 2-x can also be identified. Next appears a 'white salt' containing less than 500 ppm of plutonium, which meets the operational criterion for LLW discard. For these salts, the pyro-oxidation process in and of itself is expected to reduce the quantities to be stored on-site by more than one-third. The pyro-oxidation of CaCl 2 /NaCl base americium extraction salts leads to oxides PuO 2 and probably AmO 2 , but the yield of concentration in the black salt is lower and the white salt cannot be discarded as LLW. During vacuum distillation, excess carbonate can dissociate and damage the efficiency of the process. Appropriate chlorine sparging at the end of the oxidation can eliminate this carbonate. (authors)

  6. Reduction of uranium and plutonium oxides by aluminum. Application to the recycling of plutonium

    International Nuclear Information System (INIS)

    Gallay, J.

    1968-01-01

    A process for treating plutonium oxide calcined at high temperatures (1000 to 2000 deg. C) with a view to recovering the metal consists in the reduction of this oxide dissolved in a mixture of aluminium, sodium and calcium fluorides by aluminium at about 1180 deg. C. The first part of the report presents the results of reduction tests carried out on the uranium oxides UO 2 and U 3 O 8 ; these are in agreement with the thermodynamic calculations of the exchange reaction at equilibrium. The second part describes the application of this method to plutonium oxides. The Pu-Al alloy obtained (60 per cent Pu) is then recycled in an aqueous medium. (author) [fr

  7. The use dynamic avalanching and fractal analysis to characterise uranium oxide powders

    International Nuclear Information System (INIS)

    Hobbs, J.W.; Rhodes, D.

    2000-01-01

    Direct thermal denitration is an attractive method of co-converting mixed-metal nitrate solutions of plutonium and uranium into oxide because of its apparent simplicity. Such benefits are often marred by the relatively poor powder quality and handling characteristics, which can be overcome by modifications to the process chemistry. To ensure that powder synthesis routes under assessment require the minimal further processing it is necessary to be able to characterise the powder fully in term of the key fundamental properties. This paper will demonstrate the use of a dynamic avalanching technique, fractal analysis and morphology to assess processing behaviour. The use of dynamic avalanching to uniquely characterise the chaotic flow properties of urania powders has proved successful and results have shown that this technique is capable of detecting small differences in processing behaviour due changes in morphologies and particle size distribution. This technique has promise for being able to provide nearly instantaneous feedback to the powder generation process being monitored (e.g. calcination, milling, mixing). The use of fractals to describe powders is an interesting characterisation tool when combined with morphological shape factors and the flow index. (authors)

  8. Development of the plutonium oxide vitrification system

    International Nuclear Information System (INIS)

    Marshall, K.M.; Marra, J.C.; Coughlin, J.T.; Calloway, T.B.; Schumacher, R.F.; Zamecnik, J.R.; Pareizs, J.M.

    1998-01-01

    Repository disposal of plutonium in a suitable, immobilized form is being considered as one option for the disposition of surplus weapons-usable plutonium. Accelerated development efforts were completed in 1997 on two potential immobilization forms to facilitate downselection to one form for continued development. The two forms studied were a crystalline ceramic based on Synroc technology and a lanthanide borosilicate (LaBS) glass. As part of the glass development program, melter design activities and component testing were completed to demonstrate the feasibility of using glass as an immobilization medium. A prototypical melter was designed and built in 1997. The melter vessel and drain tube were constructed of a Pt/Rh alloy. Separate induction systems were used to heat the vessel and drain tube. A Pt/Rh stirrer was incorporated into the design to facilitate homogenization of the melt. Integrated powder feeding and off-gas systems completed the overall design. Concurrent with the design efforts, testing was conducted using a plutonium surrogate LaBS composition in an existing (near-scale) melter to demonstrate the feasibility of processing the LaBS glass on a production scale. Additionally, the drain tube configuration was successfully tested using a plutonium surrogate LaBS glass

  9. Characterization and storage of the Rocky Flats plutonium oxide legacy

    International Nuclear Information System (INIS)

    Stakebake, J.L.

    1997-01-01

    Prior to 1989, plutonium oxide storage at the U.S. Department of Energy (DOE) Rocky Flats Environmental Technology Site (RFETS) could generally be considered as short term. Packaging configurations for short-term storage consisted of slip-lid cans and sealed produce cans. Prior to packaging, most of the oxide had been stabilized by heating to ∼500 degrees C. When storage times were relatively short, few problems were attributed to either packaging or storage. However, with projected storage times extended up to 50 yr, most of the old packaging of RFETS plutonium oxide will be in two welded steel containers in compliance with the DOE Standard 3013-96, which defines processing and packaging criteria for safe, long-term storage of plutonium oxide

  10. International collaborations about fuel studies for reactor recycling of military quality plutonium

    International Nuclear Information System (INIS)

    Bernard, H.; Chaudat, J.P.

    1997-01-01

    In November 1992, an agreement was signed between the French and Russian governments to use in Russia and for pacific purposes the plutonium recovered from the Russian nuclear weapons dismantling. This plutonium will be transformed into mixed oxide fuels (MOX) for nuclear power production. The French Direction of Military Applications (DAM) of the CEA is the operator of the French-Russian AIDA program. The CEA Direction of Fuel Cycle (DCC) and Direction of Nuclear Reactors (DRN) are involved in the transformation of metallic plutonium into sinterable oxide powder for MOX fuel manufacturing. The Russian TOMOX (Treatment of MOX powder Metallic Objects) and DEMOX (MOX Demonstration) plants will produce the MOX fuel assemblies for the 4 VVER 1000 reactors of Balakovo and the fast BN 600 reactor. The second part of the program will involve the German Siemens and GRS companies for the safety studies of the reactors and fuel cycle plants. The paper gives also a brief analysis of the US policy concerning the military plutonium recycling. (J.S.)

  11. Standard test methods for chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade plutonium dioxide powders and pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 These test methods cover procedures for the chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade plutonium dioxide powders and pellets to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Plutonium Sample Handling 8 to 10 Plutonium by Controlled-Potential Coulometry Plutonium by Ceric Sulfate Titration Plutonium by Amperometric Titration with Iron(II) Plutonium by Diode Array Spectrophotometry Nitrogen by Distillation Spectrophotometry Using Nessler Reagent 11 to 18 Carbon (Total) by Direct Combustion–Thermal Conductivity 19 to 30 Total Chlorine and Fluorine by Pyrohydrolysis 31 to 38 Sulfur by Distillation Spectrophotometry 39 to 47 Plutonium Isotopic Analysis by Mass Spectrometry Rare Earth Elements by Spectroscopy 48 to 55 Trace Elements by Carrier–Distillation Spectroscopy 56 to 63 Impurities by ICP-AES Impurity Elements by Spark-Source Mass Spectrography 64 to 70 Moisture by the Coulomet...

  12. Method of manufacturing mixed stock powders for nuclear fuel elements

    International Nuclear Information System (INIS)

    Hirayama, Satoshi.

    1980-01-01

    Purpose: To alleviate the limit of the present reactor operating conditions by uniformly mixing an additive to the main content as an uranium dioxide or mixture of the uranium dioxide with plutonium dioxide. Method: A mixed stock powder is obtained by adding an additive of at least two of aluminium oxide, beryllium oxide, calcium oxide, magnesium oxide, silicon oxide, sodium oxide, potassium oxide, phosphorus oxide, titanium oxide and iron oxide to suspension having ammonia water as dispersion medium to start the deposition of precipitation at a step of precipitating ammonium diuranate or plutionium hydroxide of a main content of uranium dioxide or mixture of uranium dioxide and plutonium dioxide and deposited precipitate is calcinated and reduced. (Yoshihara, H.)

  13. Method of processing plutonium and uranium solution

    International Nuclear Information System (INIS)

    Otsuka, Katsuyuki; Kondo, Isao; Suzuki, Toru.

    1989-01-01

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

  14. LITERATURE REVIEW FOR OXALATE OXIDATION PROCESSES AND PLUTONIUM OXALATE SOLUBILITY

    Energy Technology Data Exchange (ETDEWEB)

    Nash, C.

    2012-02-03

    A literature review of oxalate oxidation processes finds that manganese(II)-catalyzed nitric acid oxidation of oxalate in precipitate filtrate is a viable and well-documented process. The process has been operated on the large scale at Savannah River in the past, including oxidation of 20 tons of oxalic acid in F-Canyon. Research data under a variety of conditions show the process to be robust. This process is recommended for oxalate destruction in H-Canyon in the upcoming program to produce feed for the MOX facility. Prevention of plutonium oxalate precipitation in filtrate can be achieved by concentrated nitric acid/ferric nitrate sequestration of oxalate. Organic complexants do not appear practical to sequester plutonium. Testing is proposed to confirm the literature and calculation findings of this review at projected operating conditions for the upcoming campaign. H Canyon plans to commence conversion of plutonium metal to low-fired plutonium oxide in 2012 for eventual use in the Mixed Oxide Fuel (MOX) Facility. The flowsheet includes sequential operations of metal dissolution, ion exchange, elution, oxalate precipitation, filtration, and calcination. All processes beyond dissolution will occur in HB-Line. The filtration step produces an aqueous filtrate that may have as much as 4 M nitric acid and 0.15 M oxalate. The oxalate needs to be removed from the stream to prevent possible downstream precipitation of residual plutonium when the solution is processed in H Canyon. In addition, sending the oxalate to the waste tank farm is undesirable. This report addresses the processing options for destroying the oxalate in existing H Canyon equipment.

  15. Simulation of uranium and plutonium oxides compounds obtained in plasma

    Science.gov (United States)

    Novoselov, Ivan Yu.; Karengin, Alexander G.; Babaev, Renat G.

    2018-03-01

    The aim of this paper is to carry out thermodynamic simulation of mixed plutonium and uranium oxides compounds obtained after plasma treatment of plutonium and uranium nitrates and to determine optimal water-salt-organic mixture composition as well as conditions for their plasma treatment (temperature, air mass fraction). Authors conclude that it needs to complete the treatment of nitric solutions in form of water-salt-organic mixtures to guarantee energy saving obtainment of oxide compounds for mixed-oxide fuel and explain the choice of chemical composition of water-salt-organic mixture. It has been confirmed that temperature of 1200 °C is optimal to practice the process. Authors have demonstrated that condensed products after plasma treatment of water-salt-organic mixture contains targeted products (uranium and plutonium oxides) and gaseous products are environmental friendly. In conclusion basic operational modes for practicing the process are showed.

  16. Screw calciner mechanical direct denitration process for plutonium nitrate to oxide conversion

    International Nuclear Information System (INIS)

    Souply, K.R.; Sperry, W.E.

    1977-01-01

    This report describes a screw calciner direct-denitration process for converting plutonium nitrate to plutonium oxide. The information should be used when making comparisons of alternative plutonium nitrate-to-oxide conversion processes or as a basis for further detailed studies. The report contains process flow sheets with a material balance; a process description; and a discussion of the process including history, advantages and disadvantages, and additional research required

  17. Oxygen potential of uranium--plutonium oxide as determined by controlled-atmosphere thermogravimetry

    International Nuclear Information System (INIS)

    Swanson, G.C.

    1975-10-01

    The oxygen-to-metal atom ratio, or O/M, of solid solution uranium-plutonium oxide reactor fuel is a measure of the concentration of crystal defects in the oxide which affect many fuel properties, particularly, fuel oxygen potential. Fabrication of a high-temperature oxygen electrode, employing an electro-active tip of oxygen-deficient solid-state electrolyte, intended to confirm gaseous oxygen potentials is described. Uranium oxide and plutonium oxide O/M reference materials were prepared by in situ oxidation of high purity metals in the thermobalance. A solid solution uranium-plutonium oxide O/M reference material was prepared by alloying the uranium and plutonium metals in a yttrium oxide crucible at 1200 0 C and oxidizing with moist He at 250 0 C. The individual and solid solution oxides were isothermally equilibrated with controlled oxygen potentials between 800 and 1300 0 C and the equilibrated O/M ratios calculated with corrections for impurities and buoyancy effects. Use of a reference oxygen potential of -100 kcal/mol to produce an O/M of 2.000 is confirmed by these results. However, because of the lengthy equilibration times required for all oxides, use of the O/M reference materials rather than a reference oxygen potential is recommended for O/M analysis methods calibrations. (auth)

  18. Conversion of metal plutonium to plutonium dioxide by pyrochemical method

    Energy Technology Data Exchange (ETDEWEB)

    Panov, A.V.; Subbotin, V.G. [Russian Federal Nuclear Center, ALL-Russian Science and Research Institute of Technical Physics, Snezhinsk (Russian Federation); Mashirev, V.P. [ALL-Russian Science and Research Institute of Chemical Technology, Moscow (Russian Federation)

    2000-07-01

    Report contains experimental results on metal plutonium of weapon origin samples conversion to plutonium dioxide by pyrochemical method. Circuits of processes are described. Their advantages and shortcomings are shown. Parameters of plutonium dioxide powders (phase and fraction compositions, poured density) manufactured by pyrochemical method in RFNC-VNIITF are shown as well. (authors)

  19. Gas generation over plutonium oxides in the 94-1 shelf-life surveillance program

    International Nuclear Information System (INIS)

    Berg, J.M.; Harradine, D.M.; Hill, D.D.; McFarlan, James T.; Padilla, D.D.; Prenger, F. Coyne; Veirs, D.K.; Worl, L.A.

    2002-01-01

    The Department of Energy (DOE) is embarking upon a program to store large quantities of plutonium-bearing materials for up to fifty years. The Los Alamos National Laboratory Shelf Life Project was established to bound the behavior of plutonium-bearing material meeting the DOE 3013 Standard. The shelf life study monitors temperature, pressure and gas composition over oxide materials in a limited number of large-scale 3013 inner containers and in many small-scale containers. For the large-scale study, baseline plutonium oxides, oxides exposed to high-humidity atmospheres, and oxides containing chloride salt impurities are planned. The first large-scale container represents a baseline and contains dry plutonium oxide prepared according to the 3013 Standard. This container has been observed for pressure, temperature and gas compositional changes for less than a year. Results indicate that no detectable changes in pressure and gas composition are observed.

  20. Development of a process for co-conversion of Pu-U nitrate mixed solutions to mixed oxide powder using microwave heating method

    International Nuclear Information System (INIS)

    Koizumi, Masumichi; Ohtsuka, Katsuyuki; Ohshima, Hirofumi; Isagawa, Hiroto; Akiyama, Hideo; Todokoro, Akio; Naruki, Kaoru

    1983-01-01

    For the complete nuclear fuel cycle, the development of a process for the co-conversion of Pu-U nitrate mixed solutions to mixed oxide powder has been performed along the line of non-proliferation policy of nuclear materials. A new co-conversion process using a microwave heating method has been developed and successfully demonstrated with good results using the test unit with a capacity of 2 kg MOX/d. Through the experiments and engineering test operations, several important data have been obtained concerning the feasibility of the test unit, powder characteristics and homogeneity of the product, and impurity pickups during denitration process. The results of these experimental operations show that the co-conversion process using a microwave heating method has many excellent advantages, such as good powder characteristics of the product, good homogeneity of Pu-U oxide, simplicity of the process, minimum liquid waste, no possibility of changing the Pu/U ratio and stable operability of the plant. Since August 1979, plutonium nitrate solution transported from the Tokai Reprocessing Plant has been converted to mixed oxide powder which has the Pu/U ratio = 1. The products have been processed to the ATR ''FUGEN'' reloading fuel. Based on the successful development of the co-conversion process, the microwave heating direct denitration facility with a 10 kg MOX/d capacity has been constructed adjacent to the reprocessing plant. This facility will come into hot operation by the fall of this year. For future development of the microwave heating method, a continuous direct denitration, a vitrification of high active liquid waste and a solidification of the plutonium-contaminated waste are investigated in Power Reactor and Nuclear Fuel Development Corp. (author)

  1. The growth and evolution of thin oxide films on delta-plutonium surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Flores, Harry G [Los Alamos National Laboratory; Pugmire, David L [Los Alamos National Laboratory

    2009-01-01

    The common oxides of plutonium are the dioxide (PuO{sub 2}) and the sesquioxide (Pu{sub 2}O{sub 3}). The structure of an oxide on plutonium metal under air at room temperature is typically described as a thick PuO{sub 2} film at the gas-oxide interface with a thinner PuO{sub 2} film near the oxide-metal substrate interface. In a reducing environment, such as ultra high vacuum, the dioxide (Pu{sup 4+}; O/Pu = 2.0) readily converts to the sesquioxide (Pu{sup 3+}; O/Pu = 1.5) with time. In this work, the growth and evolution of thin plutonium oxide films is studied with x-ray photoelectron spectroscopy (XPS) under varying conditions. The results indicate that, like the dioxide, the sesquioxide is not stable on a very clean metal substrate under reducing conditions, resulting in substoichiometric films (Pu{sub 2}O{sub 3-y}). The Pu{sub 2}O{sub 3-y} films prepared exhibit a variety of stoichiometries (y = 0.2-1) as a function of preparation conditions, highlighting the fact that caution must be exercised when studying plutonium oxide surfaces under these conditions and interpreting resulting data.

  2. A review of plutonium oxalate decomposition reactions and effects of decomposition temperature on the surface area of the plutonium dioxide product

    Science.gov (United States)

    Orr, R. M.; Sims, H. E.; Taylor, R. J.

    2015-10-01

    Plutonium (IV) and (III) ions in nitric acid solution readily form insoluble precipitates with oxalic acid. The plutonium oxalates are then easily thermally decomposed to form plutonium dioxide powder. This simple process forms the basis of current industrial conversion or 'finishing' processes that are used in commercial scale reprocessing plants. It is also widely used in analytical or laboratory scale operations and for waste residues treatment. However, the mechanisms of the thermal decompositions in both air and inert atmospheres have been the subject of various studies over several decades. The nature of intermediate phases is of fundamental interest whilst understanding the evolution of gases at different temperatures is relevant to process control. The thermal decomposition is also used to control a number of powder properties of the PuO2 product that are important to either long term storage or mixed oxide fuel manufacturing. These properties are the surface area, residual carbon impurities and adsorbed volatile species whereas the morphology and particle size distribution are functions of the precipitation process. Available data and experience regarding the thermal and radiation-induced decompositions of plutonium oxalate to oxide are reviewed. The mechanisms of the thermal decompositions are considered with a particular focus on the likely redox chemistry involved. Also, whilst it is well known that the surface area is dependent on calcination temperature, there is a wide variation in the published data and so new correlations have been derived. Better understanding of plutonium (III) and (IV) oxalate decompositions will assist the development of more proliferation resistant actinide co-conversion processes that are needed for advanced reprocessing in future closed nuclear fuel cycles.

  3. Conceptual Design for the Pilot-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Lumetta, Gregg J.; Meier, David E.; Tingey, Joel M.; Casella, Amanda J.; Delegard, Calvin H.; Edwards, Matthew K.; Jones, Susan A.; Rapko, Brian M.

    2014-08-05

    This report describes a conceptual design for a pilot-scale capability to produce plutonium oxide for use as exercise and reference materials, and for use in identifying and validating nuclear forensics signatures associated with plutonium production. This capability is referred to as the Pilot-scale Plutonium oxide Processing Unit (P3U), and it will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including plutonium dioxide (PuO2) dissolution, purification of the Pu by ion exchange, precipitation, and conversion to oxide by calcination.

  4. Plutonium chemistry: a synthesis of experimental data and a quantitative model for plutonium oxide solubility

    International Nuclear Information System (INIS)

    Haschke, J.M.; Oversby, V.M.

    2002-01-01

    The chemistry of plutonium is important for assessing potential behavior of radioactive waste under conditions of geologic disposal. This paper reviews experimental data on dissolution of plutonium oxide solids, describes a hybrid kinetic-equilibrium model for predicting steady-state Pu concentrations, and compares laboratory results with predicted Pu concentrations and oxidation-state distributions. The model is based on oxidation of PuO 2 by water to produce PuO 2+x , an oxide that can release Pu(V) to solution. Kinetic relationships between formation of PuO 2+x , dissolution of Pu(V), disproportionation of Pu(V) to Pu(IV) and Pu(VI), and reduction of Pu(VI) are given and used in model calculations. Data from tests of pyrochemical salt wastes in brines are discussed and interpreted using the conceptual model. Essential data for quantitative modeling at conditions relevant to nuclear waste repositories are identified and laboratory experiments to determine rate constants for use in the model are discussed

  5. Plutonium spot of mixed oxide fuel, 2

    International Nuclear Information System (INIS)

    Suzuki, Yukio; Maruishi, Yoshihiro; Satoh, Masaichi; Aoki, Toshimasa; Muto, Tadashi

    1974-01-01

    In a fast reactor, the specification for the homogeneity of plutonium in plutonium-uranium mixed-oxide fuel is mainly dependent on the nuclear characteristics, whereas in a thermal reactor, on thermal characteristics. This homogeneity is measured by autoradiography as the plutonium spot size of the specimens which are arbitrarily chosen fuel pellets from a lot. Although this is a kind of random sampling, it is difficult to apply this method to conventional digital standards including JIS standards. So a special sampling inspection method was studied. First, it is assumed that the shape of plutonium spots is spherical, the size distribution is logarithmic normal, and the standard deviation is constant. Then, if standard deviation and mean spot size are given, the logarithmic normal distribution is decided unitarily, and further if the total weight of plutonium spots for a lot of pellets is known, the number of the spots (No) which does not conform to the specification can be obtained. Then, the fraction defective is defined as No devided by the number of pellets per lot. As to the lot with such fraction defective, the acceptance coefficient of the lot was obtained through calculation, in which the number of sampling, acceptable diameter limit observed and acceptable conditions were used as parameters. (Tai, I.)

  6. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-07-01

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

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

    International Nuclear Information System (INIS)

    Gorbunov, V.F.; Novoselov, G.P.; Ulanov, S.A.

    1976-01-01

    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 UO 2 -PuO 2 . 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

  9. Preparation and utilization of metal oxide fine powder

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joon Soo; Jang, Hee Dong; Lim, Young Woong; Kim, Sung Don; Lee, Hi Sun; Lee, Hoo In; Kim, Chul Joo; Shim, Gun Joo; Jang, Dae Kyu [Korea Inst. of Geology Mining and Materials, Taejon (Korea, Republic of)

    1995-12-01

    Metal oxide fine powders finds many applications in industry as new materials. It is very much necessary for the development of such powders to improve the domestic industry. The purpose of present research is to develop a process for the preparation and utilization of metal oxide fine powder. This project is consisted of two main subjects. (1) Production of ultrafine metal oxide powder: Ultrafine metal oxide powder is defined as a metal oxide powder of less than 100 nanometer in particle size. Experiments for the control of particle size and distributions in the various reaction system and compared with results of (2 nd year research). Various reaction systems were adopted for the development of feasible process. Ultrafine particles could be prepared even higher concentration of TiCl{sub 4} and lower gas flowrate compared to TiCl{sub 4}-O{sub 2} system in the TiCl{sub 4}-Air-H{sub 2}O system. Ultrafine Al{sub 2}O{sub 3} powders also prepared with the change of concentration and gas flowrate. Experiments on the treatment of surface characteristics of ultrafine TiO{sub 2} powders were investigated using esterification and surface treating agents. A mathematical model that can predict the particle size and distribution was also developed. (2) Preparation of cerium oxide for high-grade polishing powder: Used cerium polishing powder was recycled for preparation of high grade cerium oxide polishing powder. Also, cerium hydroxide which was generated as by-product in processing of monazite ore was used as another material. These two materials were leached respectively by using acid, and the precipitate was gained in each leached solution by adjusting pH of the solution, and by selective crystallization. These precipitates were calcined to make high grade cerium oxide polishing powder. The effect of several experimental variables were investigated, and the optimum conditions were obtained through the experiments. (author). 81 refs., 49 figs., 27 tabs.

  10. Contamination smoke: a simulation of heavy metal containing aerosols from fires in plutonium glove boxes: part II

    International Nuclear Information System (INIS)

    Buijs, K.; Chavane de Dalmassy, B.; Baumgaertner, E.

    1992-01-01

    The study of the dispersion of plutonium bearing aerosols during glove box fires on a laboratory scale has been, in part I of this work, focussed on fires of polymethylmethacrylate (PMMA - the major glove box construction material) whose surfaces were contaminated with cerium-europium oxide powder as a substitute for plutonium-uranium oxide. The present part II completes the study with comparative fire experiments involving contaminated samples of various glove box materials burning in or exposed to the flames of the standardized 0.6 MW fire source previously developed. Beyond spreading of the Ce-Eu-oxide powder as mentioned above, the other important surface contamination process is used, i.e. deposition and subsequent drying of droplets from acid cerium-europium solutions. It is shown that, among the tested materials, and with the exception of synthetic glove rubber, burning PMMA spreads the most radioactive contamination. On the other hand, this potential risk is much lower for fires involving materials contaminated from solution deposition than from powder or pellets. Attempts to measure the airborne contaminant particle sizes did not yield conclusive results. They suggest, however, that contamination from solutions leads to smaller heavy-metal containing aerosol particles than contamination with powder

  11. Oxidation behaviour of plutonium rich (U, Pu)C and (U, Pu)O{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Sali, S.K., E-mail: sksali@barc.gov.in [Fuel Chemistry Division (India); Kulkarni, N.K.; Phatak, Rohan [Fuel Chemistry Division (India); Agarwal, Renu [Product Development Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2016-10-15

    Oxidation behaviour of (U{sub 0.3}Pu{sub 0.7})C{sub 1.06} was investigated in air by heating samples up to 1073 K and 1273 K. Thermogravimetry (TG) of the samples and X-ray powder diffraction (XRD) of the intermediate products were used to understand the phenomenon taking place during this process. Theoretical calculations were carried out to understand the multiple phase changes taking place during oxidation of carbide. Theoretical results were validated by XRD analysis of the products obtained at different stages of oxidation. The final oxidized products were found to be a single FCC phase with O/M = 2.15 (M = U + Pu). Oxidation kinetic studies of (U{sub 0.3}Pu{sub 0.7})O{sub 2} and (U{sub 0.47}Pu{sub 0.53})O{sub 2} were carried out in dry air, using thermogravimetry, under non-isothermal conditions. The activation energy of oxidation was found to be 49 and 70 kJ/mol, respectively. Lattice parameter dependence on Pu/M and O/M of plutonium rich mixed oxide (MOX) was established using combined results of XRD and TG analysis of (U{sub 0.3}Pu{sub 0.7})O{sub 2+x} and (U{sub 0.47}Pu{sub 0.53})O{sub 2+x}.

  12. 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 inhalation result in similar risk for development of lung tumors as compared with industrial plutonium oxide.

  13. An ab initio study of plutonium oxides surfaces

    International Nuclear Information System (INIS)

    Jomard, G.; Bottin, F.; Amadon, B.

    2007-01-01

    By means of first-principles calculations, we have studied the atomic structure as well as the thermodynamic stability of various plutonium dioxide surfaces in function of their environment (in terms of oxygen partial pressure and temperature). All these simulations have been performed with the ABINIT code. It is well known that DFT fails to describe correctly plutonium-based materials since 5f electrons in such systems are strongly correlated. In order to go beyond DFT, we have treated PuO 2 and β-Pu 2 O 3 in a DFT+U framework. We show that the couple of parameters (U,J) that works well for pure Pu is also well designed for describing ground state (GS) properties of these two oxides. The major improvement with respect with DFT is that we are able to predict an insulating GS in agreement with experiments. The presence of a gap in the DOS (Density of States) of plutonium oxides should play a significant role in the predicted surface reactivity. However, performing DFT+U calculations on surfaces of plutonium oxide from scratch was too ambitious. That is why we decided, as a first step, to study the stability of the (100), (110) and (111) surfaces of PuO 2 in a DFT-GGA framework. For each of these orientations, we considered various terminations. These ab initio results have been introduced in a thermodynamic model which allows us to predict the relative stability of the different terminations as a function of temperature and oxygen partial pressure (p O 2 ). We conclude that at room temperature and for p O 2 ∼10 atm., the polar O 2 -(100) termination is favoured. The stabilization of such a polar stoichiometric surface is surprising and should be confirmed by DFT+U calculations before any final conclusion. (authors)

  14. A conceptual and calculational model for gas formation from impure calcined plutonium oxides

    International Nuclear Information System (INIS)

    Lyman, John L.; Eller, P. Gary

    2000-01-01

    Safe transport and storage of pure and impure plutonium oxides requires an understanding of processes that may generate or consume gases in a confined storage vessel. We have formulated conceptual and calculational models for gas formation from calcined materials. The conceptual model for impure calcined plutonium oxides is based on the data collected to date

  15. MIS High-Purity Plutonium Oxide Metal Oxidation Product TS707001 (SSR123): Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Veirs, Douglas Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stroud, Mary Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Berg, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Narlesky, Joshua Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Worl, Laura Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Max A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Carillo, Alex [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-09

    A high-purity plutonium dioxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied with regard to gas generation and corrosion in a storage environment. Sample TS707001 represents process plutonium oxides from several metal oxidation operations as well as impure and scrap plutonium from Hanford that are currently stored in 3013 containers. After calcination to 950°C, the material contained 86.98% plutonium with no major impurities. This study followed over time, the gas pressure of a sample with nominally 0.5 wt% water in a sealed container with an internal volume scaled to 1/500th of the volume of a 3013 container. Gas compositions were measured periodically over a six year period. The maximum observed gas pressure was 138 kPa. The increase over the initial pressure of 80 kPa was primarily due to generation of nitrogen and carbon dioxide gas in the first six months. Hydrogen and oxygen were minor components of the headspace gas. At the completion of the study, the internal components of the sealed container showed signs of corrosion, including pitting.

  16. Metallographic preparation of sintered oxides, carbides and nitrides of uranium and plutonium

    International Nuclear Information System (INIS)

    Martin, A.; Arles, L.

    1967-12-01

    We describe the methods of polishing, attack and coloring used at the section of plutonium base ceramics studies. These methods have stood the test of experience on the uranium and plutonium carbides, nitrides and carbonitrides as well on the mixed uranium and plutonium oxides. These methods have been particularly adapted to fit to the low dense and sintered samples [fr

  17. Oxidation of nano-sized aluminum powders

    International Nuclear Information System (INIS)

    Vorozhtsov, A.B.; Lerner, M.; Rodkevich, N.; Nie, H.; Abraham, A.; Schoenitz, M.; Dreizin, E.L.

    2016-01-01

    Highlights: • Weight gain measured in TG oxidation experiments was split between particles of different sizes. • Reaction kinetics obtained by isoconversion explicitly accounting for the effect of size distribution. • Activation energy is obtained as a function of oxide thickness for growth of amorphous alumina. • Oxidation mechanism for nanopowders remains the same as for coarser aluminum powders. - Abstract: Oxidation of aluminum nanopowders obtained by electro-exploded wires is studied. Particle size distributions are obtained from transmission electron microscopy (TEM) images. Thermo-gravimetric (TG) experiments are complemented by TEM and XRD studies of partially oxidized particles. Qualitatively, oxidation follows the mechanism developed for coarser aluminum powder and resulting in formation of hollow oxide shells. Sintering of particles is also observed. The TG results are processed to account explicitly for the particle size distribution and spherical shapes, so that oxidation of particles of different sizes is characterized. The apparent activation energy is obtained as a function of the reaction progress using model-free isoconversion processing of experimental data. A complete phenomenological oxidation model is then proposed assuming a spherically symmetric geometry. The oxidation kinetics of aluminum powder is shown to be unaffected by particle sizes reduced down to tens of nm. The apparent activation energy describing growth of amorphous alumina is increasing at the very early stages of oxidation. The higher activation energy is likely associated with an increasing homogeneity in the growing amorphous oxide layer, initially containing multiple defects and imperfections. The trends describing changes in both activation energy and pre-exponent of the growing amorphous oxide are useful for predicting ignition delays of aluminum particles. The kinetic trends describing activation energies and pre-exponents in a broader range of the oxide

  18. Plutonium oxides analysis. Sulfur potentiometric analysis

    International Nuclear Information System (INIS)

    Anon.

    Total sulfur determination (sulfur, sulfates, sulfides ...) in plutonium oxides, suitable for sulfate ion content between 0.003 percent to 0.2 percent, by dissolution in nitric hydrofluoric acid, nitrates elimination, addition of hydrochloric acid and reduction in hydrogen sulfide which is carried by an inert gas and neutralized by sodium hydroxide. Sodium sulfide is titrated with mercuric acetate by constant intensity potentiometry [fr

  19. Dynamic process model of a plutonium oxalate precipitator. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.L.; Hammelman, J.E.; Borgonovi, G.M.

    1977-11-01

    In support of LLL material safeguards program, a dynamic process model was developed which simulates the performance of a plutonium (IV) oxalate precipitator. The plutonium oxalate precipitator is a component in the plutonium oxalate process for making plutonium oxide powder from plutonium nitrate. The model is based on state-of-the-art crystallization descriptive equations, the parameters of which are quantified through the use of batch experimental data. The dynamic model predicts performance very similar to general Hanford oxalate process experience. The utilization of such a process model in an actual plant operation could promote both process control and material safeguards control by serving as a baseline predictor which could give early warning of process upsets or material diversion. The model has been incorporated into a FORTRAN computer program and is also compatible with the DYNSYS 2 computer code which is being used at LLL for process modeling efforts.

  20. Dynamic process model of a plutonium oxalate precipitator. Final report

    International Nuclear Information System (INIS)

    Miller, C.L.; Hammelman, J.E.; Borgonovi, G.M.

    1977-11-01

    In support of LLL material safeguards program, a dynamic process model was developed which simulates the performance of a plutonium (IV) oxalate precipitator. The plutonium oxalate precipitator is a component in the plutonium oxalate process for making plutonium oxide powder from plutonium nitrate. The model is based on state-of-the-art crystallization descriptive equations, the parameters of which are quantified through the use of batch experimental data. The dynamic model predicts performance very similar to general Hanford oxalate process experience. The utilization of such a process model in an actual plant operation could promote both process control and material safeguards control by serving as a baseline predictor which could give early warning of process upsets or material diversion. The model has been incorporated into a FORTRAN computer program and is also compatible with the DYNSYS 2 computer code which is being used at LLL for process modeling efforts

  1. Recombination rates of hydrogen and oxygen over pure and impure plutonium oxides

    International Nuclear Information System (INIS)

    Morales, L.

    1999-01-01

    Long-term, safe storage of excess plutonium-bearing materials is required until stabilization and disposal methods are implemented or defined. The US Department of Energy (DOE) has established a plan to address the stabilization, packing, and storage of plutonium-bearing materials from around the complex. The DOE's standard method, DOE-STD-3013-96 and its proposed revision, for stabilizing pure and impure actinide materials is by calcination in air followed by sealing the material in welded stainless steel containers. The 3013 standard contains and equation that predicts the total pressure buildup in the can over the anticipated storage time of 50 yr. This equation was meant to model a worst-case scenario to ensure that pressures would not exceed the strength of the container at the end of 50 yr. As a result, concerns about pressure generation in the storage cans, both absolute values and rates, have been raised with regard to rupture and dispersal of nuclear materials. Similar issues have been raised about the transportation of these materials around the complex. The purpose of this work is to provide a stronger technical basis for the 3013 standard by measuring the recombination rates of hydrogen/oxygen mixtures in contact with pure and impure plutonium oxides. The goal of these experiments was to determine whether the rate of recombination is faster than the rate of water radiolysis under controlled conditions. This was accomplished by using a calibrated pressure-volume-temperature apparatus to measure the recombination rates in a fixed volume as the gas mixture was brought into contact with oxide powders whose temperatures ranged from 50 to 300 C. These conditions were selected in order to bracket the temperature conditions expected in a typical storage can. In addition, a 2% H 2 /air mixture encompasses scenarios in which the cans are sealed in air, and over time various amounts of hydrogen are formed

  2. High-temperature vacuum distillation separation of plutonium waste salts

    International Nuclear Information System (INIS)

    Garcia, E.

    1996-01-01

    In this task, high-temperature vacuum distillation separation is being developed for residue sodium chloride-potassium chloride salts resulting from past pyrochemical processing of plutonium. This process has the potential of providing clean separation of the salt and the actinides with minimal amounts of secondary waste generation. The process could produce chloride salt that could be discarded as low-level waste (LLW) or low actinide content transuranic (TRU) waste, and a concentrated actinide oxide powder that would meet long-term storage standards (DOE-DTD-3013-94) until a final disposition option for all surplus plutonium is chosen

  3. Determination of plutonium in pure plutonium nitrate solutions - Gravimetric method

    International Nuclear Information System (INIS)

    1987-01-01

    This International Standard specifies a precise and accurate gravimetric method for determining the concentration of plutonium in pure plutonium nitrate solutions and reference solutions, containing between 100 and 300 g of plutonium per litre, in a nitric acid medium. The weighed portion of the plutonium nitrate is treated with sulfuric acid and evaporated to dryness. The plutonium sulfate is decomposed and formed to oxide by heating in air. The oxide is ignited in air at 1200 to 1250 deg. C and weighed as stoichiometric plutonium dioxide, which is stable and non-hygroscopic

  4. The Plutonium Fuel Laboratory at Studsvik and Its Activities

    Energy Technology Data Exchange (ETDEWEB)

    Hultgren, A.; Berggren, G.; Brown, A.; Eng, H. U.; Forsyth, R. S. [AB Atomenergi, Studsvik (Sweden)

    1967-09-15

    The plutonium fuel laboratory at Studsvik is engaged in development work on plutonium-enriched fuel. At present, low enriched fuel for thermal reactors is being studied: work on fuel with a higher plutonium content for fast reactors is foreseen at a later date. So far only the pellet technique is under consideration, and a number of pellet rod specimens will be produced and irradiated in the reactor R2. These specimens include pellets from both co-precipitated uranium-plutonium salts and from physically mixed oxides. Comparison of these two materials will be extended to different density levels and different heat ratings. The methods and techniques used and studied include wet chemical work for powder preparation (continuous precipitation of Pu(IV)-oxalate with oxalic acid, continuous co-precipitation of plutonium and uranium with ammonia, optimization of.precipitation conditions using U(IV) and U(VI) respectively) ; powder preparation (drying, calcination, reduction, mixing, milling, binder addition, granulation); pellet preparation (pressing, debonding, sintering, inspection): encapsulation (charging, welding of end plug, helium filling, end sealing by welding, leak detection, decontamination); metallography (specimen preparation (moulding, polishing), etching, microscopy); structure investigations (thermal analysis (TG, DTA), X-ray diffraction, neutron diffraction, data handling by computer analysis); radiometric methods (direct plutonium determination by gamma spectrometry, non-destructive burn-up analysis by high resolution gamma spectrometry, using a Ge(Li) detector) ; rework of waste (recovery of plutonium from fuel waste by extraction with trilauryl amine and anion exchange). The plutonium fuel laboratory forms part of the Active Central Laboratory. The equipment is contained in four adjacent 10 x 15 m rooms; .for diffraction work and inactive uranium work additional space is available. All the forty glove boxes in operation except two are of AB Atomenergi

  5. CONVERSION OF PLUTONIUM TRIFLUORIDE TO PLUTONIUM TETRAFLUORIDE

    Science.gov (United States)

    Fried, S.; Davidson, N.R.

    1957-09-10

    A large proportion of the trifluoride of plutonium can be converted, in the absence of hydrogen fluoride, to the tetrafiuoride of plutonium. This is done by heating plutonium trifluoride with oxygen at temperatures between 250 and 900 deg C. The trifiuoride of plutonium reacts with oxygen to form plutonium tetrafluoride and plutonium oxide, in a ratio of about 3 to 1. In the presence of moisture, plutonium tetrafluoride tends to hydrolyze at elevated temperatures and therefore it is desirable to have the process take place under anhydrous conditions.

  6. Fluid bed direct denitration process for plutonium nitrate to oxide conversion

    International Nuclear Information System (INIS)

    Souply, K.R.; Neal, D.H.

    1977-01-01

    The fluid bed direct-denitration process appears feasible for reprocessing Light Water Reactor fuel. Considerable experience with the fluid bed process exists in the denitration of uranyl nitrate and it shows promise for use in the denitration of plutonium nitrate. The process will require some development work before it can be used in a production-size facility. This report describes a fluid bed direct-denitration process for converting plutonium nitrate to plutonium oxide, and the information should be used when making comparisons of alternative processes or as a basis for further detailed studies

  7. A case of internal contamination with plutonium oxide

    International Nuclear Information System (INIS)

    Breuer, F.; Clemente, G.F.; Strambi, E.; Testa, C.

    1980-01-01

    A case is described of internal contamination caused by the accidental explosion of a glove box, whereby a technician was wounded in his right cheek and inhaled insoluble PuO 2 . Immediate washing of the wound with a DTPA solution and the performance of a small surgical toilet reduced the wound contamination a hundred times. Intravenous injection of DPTA was started and continued for nine days. The ''in vivo'' determination of plutonium and Am-241 lung content was performed immediately after the accident and several other times thereafter. Plutonium was determined periodically in urine and fecal samples for five months. The large Esub(f)/Esub(u) ratio and the steep slope of the Esub(f) curve in the first days indicated that the inhaled material was very insoluble and that the suspended powder particles were large. This conclusion was confirmed by the lung clearance ''in vivo'' during the first few days. The initial lung burden, calculated by the fecal excretion of plutonium in the first five days, was in very good agreement with the direct measurement of the lung content; the residual lung burden and the systemic burden were derived respectively U by the fecal and urinary excretion 100 days after the intake. The committed doses to lungs, bone, liver and kidneys were calculated and were found to be considerably lower than the maximum permissible levels. (H.K.)

  8. Literature review for oxalate oxidation processes and plutonium oxalate solubility

    Energy Technology Data Exchange (ETDEWEB)

    Nash, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-01

    A literature review of oxalate oxidation processes finds that manganese(II)-catalyzed nitric acid oxidation of oxalate in precipitate filtrate is a viable and well-documented process. The process has been operated on the large scale at Savannah River in the past, including oxidation of 20 tons of oxalic acid in F-Canyon. Research data under a variety of conditions show the process to be robust. This process is recommended for oxalate destruction in H-Canyon in the upcoming program to produce feed for the MOX facility. Prevention of plutonium oxalate precipitation in filtrate can be achieved by concentrated nitric acid/ferric nitrate sequestration of oxalate. Organic complexants do not appear practical to sequester plutonium. Testing is proposed to confirm the literature and calculation findings of this review at projected operating conditions for the upcoming campaign.

  9. Supporting information for the estimation of plutonium oxide leak rates through very small apertures

    International Nuclear Information System (INIS)

    Schwendiman, L.C.

    1977-01-01

    Information is presented from which an estimate can be made of the release of plutonium oxide from shipping containers. The leak diameter is estimated from gas leak tests of the container and an estimate is made of gas leak rate as a function of pressure over the time of interest in the accident. These calculations are limited in accuracy because of assumptions regarding leak geometry and the basic formulations of hydrodynamic flow for the assumed conditions. Sonic flow is assumed to be the limiting gas flow rate. Particles leaking from the air space above the powder will be limited by the low availability of particles due to rapid settling, the very limited driving force (pressure buildup) during the first minute, and the deposition in the leak channel. Equations are given to estimate deposition losses. Leaks of particles occurring below the level of the bulk powder will be limited by mechanical interference when leaks are of dimension smaller than particle sizes present. Some limiting cases can be calculated. When the leak dimension is large compared to the particle sizes present, maximum particle releases can be estimated, but will be very conservative

  10. Preconceptual design for separation of plutonium and gallium by ion exchange

    International Nuclear Information System (INIS)

    DeMuth, S.F.

    1997-01-01

    The disposition of plutonium from decommissioned nuclear weapons, by incorporation into commercial UO 2 -based nuclear reactor fuel, is a viable means to reduce the potential for theft of excess plutonium. This fuel, which would be a combination of plutonium oxide and uranium oxide, is referred to as a mixed oxide (MOX). Following power generation in commercial reactors with this fuel, the remaining plutonium would become mixed with highly radioactive fission products in a spent fuel assembly. The radioactivity, complex chemical composition, and large size of this spent fuel assembly, would make theft difficult with elaborate chemical processing required for plutonium recovery. In fabricating the MOX fuel, it is important to maintain current commercial fuel purity specifications. While impurities from the weapons plutonium may or may not have a detrimental affect on the fuel fabrication or fuel/cladding performance, certifying the effect as insignificant could be more costly than purification. Two primary concerns have been raised with regard to the gallium impurity: (1) gallium vaporization during fuel sintering may adversely affect the MOX fuel fabrication process, and (2) gallium vaporization during reactor operation may adversely affect the fuel cladding performance. Consequently, processes for the separation of plutonium from gallium are currently being developed and/or designed. In particular, two separation processes are being considered: (1) a developmental, potentially lower cost and lower waste, thermal vaporization process following PuO 2 powder preparation, and (2) an off-the-shelf, potentially higher cost and higher waste, aqueous-based ion exchange (IX) process. While it is planned to use the thermal vaporization process should its development prove successful, IX has been recommended as a backup process. This report presents a preconceptual design with material balances for separation of plutonium from gallium by IX

  11. The use of the average plutonium-content for criticality evaluation of boiling water reactor mixed oxide-fuel transport and storage packages

    International Nuclear Information System (INIS)

    Mattera, C.

    2003-01-01

    Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and (Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by an homogenous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, Cogema Logistics (formerly Transnucleaire) has studied a new calculation method based on the use of the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in the paper. (author)

  12. A Calibration to Predict Concentrations of Impurities in Plutonium Oxide by Prompt Gamma Analysis

    International Nuclear Information System (INIS)

    Narlesky, J.E.; Kelly, E.J.; Foster, L.A.

    2005-01-01

    Prompt gamma (PG) analysis has been used to identify the presence of certain impurities in plutonium oxide, which has been stored in 3013 containers. A regression analysis was used to evaluate the trends between the count rates obtained from PG analysis and the concentration of the impurities in plutonium oxide samples measured by analytical chemistry techniques. The results of the analysis were used to obtain calibration curves, which may be used to predict the concentration of Al, Be, Cl, F, Mg, and Na in the 3013 containers. The scatter observed in the data resulted from several factors including sample geometry, error in sampling for chemical assay, statistical counting error, and intimacy of mixing of impurities and plutonium. Standards prepared by mixing plutonium oxide with CaF 2 , NaCl, and KCl show that intimacy mixing and sampling error have the largest influence on the results. Although these factors are difficult to control, the calibrations are expected to yield semiquantitative results that are sufficient for the purpose of ordering or ranking

  13. Preparation of cerium oxide for lens polishing powder

    International Nuclear Information System (INIS)

    Injarean, Uthaiwan; Rodthongkom, Chouvana; Pichestapong, Pipat; Changkrurng, Kalaya

    2003-10-01

    Cerium is an element of rare earth group which is called lanthanide series. It is found in the ores like monazite and xenotime which are the tailings of tin mines in the south of Thailand. Cerium is used mostly as lens polishing powder besides the applications in other industries. In this study, cerium extracted from monazite ore breakdown by alkaline process was used for the preparation of lens polishing powder. Cerium hydroxide cake from the process was dissolved by hydrochloric acid and precipitated with oxalic acid. The oxalate precipitate then was calcined to oxide powder and its particle size was measured. Precipitation conditions being studied are concentration of feed cerium chloride solution, concentration of oxalic acid used for the precipitation, concentration of sulfuric acid used as precipitation control reagent and the precipitation temperature. It was found that the appropriate precipitation conditions yielded the fine oxide powder with particle size about 12μm. The oxide powder can be ground to the size of 1-3 μm which is suitable for making lens polishing powder

  14. Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication

    International Nuclear Information System (INIS)

    Hanus, D.; Kleykamp, H.

    1982-01-01

    Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)

  15. Confirmation test of powder mixing process in J-MOX

    International Nuclear Information System (INIS)

    Ota, Hiroshi; Osaka, Shuichi; Kurita, Ichiro

    2009-01-01

    Japan Nuclear Fuel Ltd. (hereafter, JNFL) MOX Fuel Fabrication Plant (hereafter, J-MOX) is what fabricates MOX fuel for domestic light water power plants. Development of design concept of J-MOX was started mid 90's and the frame of J-MOX process was clarified around 2000 including adoption of MIMAS process as apart of J-MOX powder process. JNFL requires to take an answer to any technical question that has not been clarified ever before by world's MOX and/or Uranium fabricators before it commissions equipment procurement. J-MOX is to be constructed adjacent to the Rokkasho Reprocessing Plant (RRP) and to utilize MH-MOX powder recovered at RRP. The combination of the MIMAS process and the MH-MOX powder is what has never tried in the world. Therefore JNFL started a series of confirmation tests of which the most important is the powder test to confirm the applicability of MH-MOX powder to the MIMAS process. The MH-MOX powder, consisting of 50% plutonium oxide and 50% uranium oxide, originates JAEA development utilizing microwave heating (MH) technology. The powder test started with laboratory scale small equipment utilizing both uranium and the MOX powder in 2000, left a solution to tough problem such as powder adhesion onto equipment, and then was followed by a large scale equipment test again with uranium and the MOX powder. For the MOX test, actual size equipment within glovebox was manufactured and installed in JAEA plutonium fuel center in 2005, and based on results taken so far an understanding that the MIMAS equipment, with the MH-MOX powder, can present almost same quality MOX pellet as what is introduced as fabricated in Europe was developed. The test was finished at the end of Japanese fiscal year (JFY) 2007, and it was confirmed that the MOX pellets fabricated in this test were almost satisfied with the targeted specifications set for domestic LWR MOX fuels. (author)

  16. Review Of Plutonium Oxidation Literature

    International Nuclear Information System (INIS)

    Korinko, P.

    2009-01-01

    A brief review of plutonium oxidation literature was conducted. The purpose of the review was to ascertain the effect of oxidation conditions on oxide morphology to support the design and operation of the PDCF direct metal oxidation (DMO) furnace. The interest in the review was due to a new furnace design that resulted in oxide characteristics that are different than those of the original furnace. Very little of the published literature is directly relevant to the DMO furnace operation, which makes assimilation of the literature data with operating conditions and data a convoluted task. The oxidation behavior can be distilled into three regimes, a low temperature regime (RT to 350 C) with a relatively slow oxidation rate that is influenced by moisture, a moderate temperature regime (350-450 C) that is temperature dependent and relies on more or less conventional oxidation growth of a partially protective oxide scale, and high temperature oxidation (> 500 C) where the metal autocatalytically combusts and oxidizes. The particle sizes obtained from these three regimes vary with the finest being from the lowest temperature. It is surmised that the slow growth rate permits significant stress levels to be achieved that help break up the oxides. The intermediate temperatures result in a fairly compact scale that is partially protective and that grows to critical thickness prior to fracturing. The growth rate in this regime may be parabolic or paralinear, depending on the oxidation time and consequently the oxide thickness. The high temperature oxidation is invariant in quiescent or nearly quiescent conditions due to gas blanketing while it accelerates with temperature under flowing conditions. The oxide morphology will generally consist of fine particles ( 250 (micro)m). The particle size ratio is expected to be < 5%, 25%, and 70% for fine, medium and large particles, respectively, for metal temperatures in the 500-600 C range.

  17. Simulation and analysis of the plutonium oxide/metal storage containers subject to various loading conditions

    International Nuclear Information System (INIS)

    Gong, C.; Miller, R.F.

    1995-05-01

    The structural and functional requirements of the Plutonium Oxide/Metal Storage Containers are specified in the Report ''Complex 21 Plutonium Storage Facility Material Containment Team Technical Data Report'' [Complex 21, 1993]. There are no existing storage containers designed for long term storage of plutonium and current codes, standards or regulations do not adequately cover this case. As there is no extensive experience with the long term (50+ years) storage of plutonium, the design of high integrity storage containers must address many technical considerations. This analysis discusses a few potential natural phenomena that could theoretically adversely affect the container integrity over time. The plutonium oxide/metal storage container consists of a primary containment vessel (the outer container), a bagless transfer can (the inner container), two vertical plates on top of the primary containment vessel, a circular plate (the flange) supported by the two plates, tube for gas sampling operations mounted at the center of the primary containment vessel top and a spring system being inserted in the cavity between the primary containment vessel and the cap of the bagless transfer can. The dimensions of the plutonium oxide/metal storage container assembly can be found in Figure 2-1. The primary container, the bagless transfer can, and all the attached components are made of Type 304L stainless steel

  18. Experiences of shipper-receiver differences in plutonium oxide transactions

    International Nuclear Information System (INIS)

    Swinburn, K.A.; McGowan, I.R.

    1976-01-01

    Plutonium oxide has the property of absorbing moisture, carbon dioxide etc., to the extent that it is necessary to take special steps to determine the plutonium content in such a way as to avoid shipper-receiver differences. This paper describes a method used internationally for several years for correcting for any absorption. The consignee witnesses the weighing of the bulk material (D kg) at the time of blending and sampling at the consignor's Works, and also the weighing of the sample (C g). Immediately before analysis the sample is reweighed (B g) and assayed for plutonium (A%). The weight of plutonium is then (A/100)x(B/C)xD kg. The accepted weight is the average between consignor's and consignee's figures after both have used the formula indicated. It goes on to give actual assay and fissile content figures obtained by BNFL and its consignees for the international transactions already referred to, and concludes that the method of approach results cumulatively in a negligible shipper-receiver difference. Individual transactions are shown to be commercially acceptable and, it is considered, also acceptable for safeguards purposes. It is suggested that for safeguarded material the safeguards inspectors might also be able to be present at, and themselves witness, the blending, sampling and weighing operations, and at the same time take their own samples for analysis. It is further suggested that the principle could be extended to accoutancy for plutonium in plutonium nitrate solutions provided leakage does not occur from the sample vials. (author)

  19. Non intrusive measurement of residual moisture rate of plutonium oxide powder in sealed boxes; Mesure non intrusive du taux d`humidite residuel de la poudre d`oxyde de plutonium contenue dans des boites scellees

    Energy Technology Data Exchange (ETDEWEB)

    Pochet, T.; Edeline, J.C.; Domenech, T.

    1993-12-31

    This document deals with the best method to create a non intrusive measurement of residual moisture of plutonium oxide in stainless steel sealed boxes (sensibility, precision, reproducibility and feasibility in laboratory). This method is the neutron spectrometry by {sup 3}He (n,p) {sup 3}He reaction. (TEC). 16 refs., 16 figs., 2 tabs.

  20. Precipitation of plutonium from acidic solutions using magnesium oxide

    International Nuclear Information System (INIS)

    Jones, S.A.

    1994-01-01

    Plutonium (IV) is only marginally soluble in alkaline solution. Precipitation of plutonium using sodium or potassium hydroxide to neutralize acidic solutions produces a gelatinous solid that is difficult to filter and an endpoint that is difficult to control. If the pH of the solution is too high, additional species precipitate producing an increased volume of solids separated. The use of magnesium oxide as a reagent has advantages. It is added as a solid (volume of liquid waste produced is minimized), the pH is self-limiting (pH does not exceed about 8.5), and the solids precipitated are more granular (larger particle size) than those produced using KOH or NaOH. Following precipitation, the raffinate is expected to meet criteria for disposal to tank farms. The solid will be heated in a furnace to dry it and convert any hydroxide salts to the oxide form. The material will be cooled in a desiccator. The material is expected to meet vault storage criteria

  1. Stop plutonium

    International Nuclear Information System (INIS)

    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.)

  2. Plutonium solubilities

    International Nuclear Information System (INIS)

    Puigdomnech, I.; Bruno, J.

    1991-02-01

    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)

  3. Weapons-grade plutonium dispositioning. Volume 4

    International Nuclear Information System (INIS)

    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 2 -ZrO 2 -CaO) with the addition of thorium oxide (ThO 2 ) or a burnable poison such as erbium oxide (Er 2 O 3 ) or europium oxide (Eu 2 O 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 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. Airborne Release of Particles in Overheating Incidents Involving Plutonium Metal and Compounds

    Energy Technology Data Exchange (ETDEWEB)

    Schwendiman, L. C.; Mishima, J.; Radasch, C. A. [Battelle Memorial Institute, Pacific Northwest Laboratory, Richland, WA (United States)

    1968-12-15

    Ever-increasing utilization of nuclear fuels will result in wide-scale plutonium recovery processing, reconstitution of fuels, transportation, and extensive handling of this material. A variety of circumstances resulting in overheating and fires involving plutonium may occur, releasing airborne particles. This work describes the observations from a study in which the airborne release of plutonium and its compounds was measured during an exposure of the material of interest containing plutonium to temperatures which may result from fires. Aerosol released from small cylinders of metallic plutonium ignited in air at temperatures from 410 to 650 Degree-Sign C ranged from 3 x 10{sup -6} to 5 x 10{sup -5} wt%. Particles smaller than 15{mu}m in diameter represented as much as 0.03% of the total released. Large plutonium pieces weighing from 456 to 1770 g were ignited and allowed to oxidize completely in air with a velocity of around 500 cm/sec. Release rates of from 0.0045 to 0.032 wt% per hour were found. The median mass diameter of airborne material was 4 {mu}m. Quenching the oxidation with magnesium oxide sand reduced the release to 2.9 X 10{sup -4} wt% per hour. Many experiments were carried out in which plutonium compounds as powders were heated at temperatures ranging from 700 to 1000 Degree-Sign C with several air flows. Release rates ranged from 5 x 10{sup -8} to 0.9 wt% per hour, depending upon the compound and the conditions imposed. The airborne release from boiling solutions of plutonium nitrate were roughly related to energy of boiling, and ranged from 4 x 10{sup -4} to 2 x 10{sup -1} % for the evaporation of 90% of the solution. The fraction airborne when combustibles contaminated with plutonium are burned is under study. The data reported can be used in assessing the consequences of off-standard situations involving plutonium and its compounds in fires. (author)

  5. Laboratory techniques for safe encapsulation of α-emitting powder samples

    International Nuclear Information System (INIS)

    Chamberlain, H.E.; Pottinger, J.S.

    1984-01-01

    Plutonium oxide powder samples can be encapsulated in thin plastic film to prevent spread of contamination in counting and X-ray diffraction equipment. The film has to be thin enough to transmit X-rays and α-particles. Techniques are described for the wrapping process and the precautions necessary to keep the sample processing line free of significant contamination. (author)

  6. Fused salt processing of impure plutonium dioxide to high-purity plutonium metal

    International Nuclear Information System (INIS)

    Mullins, L.J.; Christensen, D.C.; Babcock, B.R.

    1982-01-01

    A process for converting impure plutonium dioxide (approx. 96% pure) to high-purity plutonium metal (>99.9%) was developed. The process consists of reducing the oxide to an impure plutonium metal intermediate with calcium metal in molten calcium chloride. The impure intermediate metal is cast into an anode and electrorefined to produce high-purity plutonium metal. The oxide reduction step is being done now on a 0.6-kg scale with the resulting yield being >99.5%. The electrorefining is being done on a 4.0-kg scale with the resulting yield being 80 to 85%. The purity of the product, which averages 99.98%, is essentially insensitive to the purity of the feed metal. The yield, however, is directly dependent on the chemical composition of the feed. To date, approximately 250 kg of impure oxide has been converted to pure metal by this processing sequence. The availability of impure plutonium dioxide, together with the need for pure plutonium metal, makes this sequence a valuable plutonium processing tool

  7. Solid-phase extraction of plutonium in various oxidation states from simulated groundwater using N-benzoylphenylhydroxylamine

    International Nuclear Information System (INIS)

    Perevalov, S.A.; Malofeeva, G.I.; Kuzovkina, E.V.; Spivakov, B.Ya.

    2013-01-01

    Solid-phase extraction of plutonium in different individual and mixed oxidation states from simulated groundwater (pH 8.5) was studied. The extraction of plutonium species was carried out in a dynamic mode using DIAPAK C16 cartridges modified by N-benzoylphenylhydroxylamine (BPHA). It was shown that the extent of recovery depends on the oxidation state of plutonium. The extraction of Pu(IV) was at the level of 98-99% regardless of the volume and flow-rate of the sample solution. Pu(V) was extracted by 90-95% and 75-80% from 10- and 100-mL aliquots of the samples, respectively, whereas the extraction of Pu(VI) did not exceed 45-50%. An equimolar mixture of Pu(IV), Pu(V), and Pu(VI) was extracted by 74%. The distribution coefficients (K d ) and kinetic exchange capacities (S) of plutonium in various oxidation states were measured. It was found that during the sorption process, Pu(V) was reduced to Pu(IV) by 80-90% after an hour-long contact with the solid phase. Pu(VI) is reduced to Pu(V) by 34% and to Pu(IV) by 55%. In the case of mixed-valent solution of plutonium, only Pu(V) and Pu(IV) were found in the effluents. (author)

  8. Process for fabricating mixed-oxide powders

    International Nuclear Information System (INIS)

    Elmaleh, D.; Giraudel, A.

    1975-01-01

    A physical-chemical process for fabricating homogeneous powders suitable for sintering is described. It can be applied to the synthesis of all mixed oxides having mutually compatible and water soluble salts. As a specific example, the fabrication of lead titanate-zirconate powders used to make hot pressed ceramics is described. These ceramics show improved piezoelectric properties [fr

  9. Preparation of hexavalent plutonium and its determination in the presence of tetravalent plutonium; Preparation de plutonium hexavalent et dosage en presence de plutonium tetravalent

    Energy Technology Data Exchange (ETDEWEB)

    Corpel, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Corpel, J [Institut du Radium, 75 - Paris (France)

    1958-07-01

    In order to study the eventual reduction of plutonium from the VI-valent state to the IV-valent state, in sulphuric medium, under the influence of its own {alpha} radiation or of the {gamma}-rays from a cobalt-60 source, we have developed a method for preparing pure hexavalent plutonium and two methods for determining solutions containing tetravalent and hexavalent plutonium simultaneously. Hexavalent plutonium was prepared by anodic oxidation at a platinum electrode. Study of the oxidation yield as a function of various factors has made it possible to define experimental conditions giving complete oxidation. For concentrations in total plutonium greater than 1.5 x 10{sup -3} M, determination of the two valencies IV and VI was carried out by spectrophotometry at two wavelengths. For lower concentrations, the determination was done by counting, after separation of the tetravalent plutonium in the form of fluoride in the presence of a carrier. (author) [French] Afin d'etudier l'eventuelle reduction du plutonium de l'etat de valence VI a l'etat de valence IV, en milieu sulfurique sous l'influence de son propre rayonnement {alpha} ou des rayons {gamma} d'une source de cobalt-60, nous avons mis au point une methode de preparation de plutonium hexavalent pur et deux methodes de dosage des solutions contenant simultanement du plutonium tetravalent et du plutonium hexavalent. Nous avons prepare le plutonium hexavalent par oxydation anodique au contact d'une electrode de platine. L'etude de rendement de l'oxydation en fonction des divers facteurs nous a permis de definir des conditions experimentales donnant une oxydation complete. Pour des concentrations en plutonium total superieures a 1,5.10{sup -3} M, le dosage des deux valences IV et VI a ete realise par spectrophotometrie a deux longueurs d'onde. Pour des concentrations inferieures, le dosage a ete effectue par comptage apres separation du plutonium tetravalent sous la forme du fluorure en presence d'un entraineur

  10. Thorium utilization as a Pu-burner: proposal of Plutonium-Thorium Mixed Oxide (PT-MOX) Project

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    2000-01-01

    In this paper, a Pu-Th mixed oxide (PT-MOX) project is proposed for a thorium utilization and a plutonium burning. None of plutonium can be newly produced from PT-MOX fuel, and the plutonium mass of about 1 ton can be consumed with one reactor (total heavy metal assumed: 100 tons) for 1 year. In order to consume plutonium produced from usual Light Water Reactor, it should be better to operate one PT-MOX reactor for three to five Light Water Reactors. (author)

  11. Fluorine and chlorine determination in mixed uranium-plutonium oxide fuel and plutonium dioxide

    International Nuclear Information System (INIS)

    Elinson, S.V.; Zemlyanukhina, N.A.; Pavlova, I.V.; Filatkina, V.P.; Tsvetkova, V.T.

    1981-01-01

    A technique of fluorine and chlorine determination in the mixed uranium-plutonium oxide fuel and plutonium dioxide, based on their simultaneous separation by means of pyrohydrolysis, is developed. Subsequently, fluorine is determined by photometry with alizarincomplexonate of lanthanum or according to the weakening of zirconium colouring with zylenol orange. Chlorine is determined using the photonephelometric method according to the reaction of chloride-ion interaction with silver nitrate or by spectrophotometric method according to the reaction with mercury rhodanide. The lower limit of fluorine determination is -6x10 -5 %, of chlorine- 1x10 -4 % in the sample of 1g. The relative mean quadratic deviation of the determination result (Ssub(r)), depends on the character of the material analyzed and at the content of nx10 -4 - nx10 -3 mass % is equal to from 0.05 to 0.32 for fluorine and from 0.11 to 0.35 for chlorine [ru

  12. Plutonium-241 processing: from impure oxide to high purity metal target disks

    International Nuclear Information System (INIS)

    Conner, W.V.; Baaso, D.L.

    1975-01-01

    The preparation of three plutonium-241 metal target disks, using a precision casting technique, is described. The disks were 0.625 inch in diameter and 0.125, 0.025, and 0.010 inch thick. All three disks were prepared simultaneously in a single casting. The variation in thickness of each disk was within +-1 percent of the disk's average thickness. The plutonium-241 was highly pure, and the finished disks contained a total of only 297 parts per million of detectable impurities. Purification of the plutonium oxide ( 241 PuO 2 ) and the conversion of the purified 241 PuO 2 to metal are also described. (U.S.)

  13. Sludge stabilization at the Plutonium Finishing Plant, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1994-10-01

    This Environmental Assessment evaluates the proposed action to operate two laboratory-size muffle furnaces in glovebox HC-21C, located in the Plutonium Finishing Plant (PFP), Hanford Site, Richland, Washington. The muffle furnaces would be used to stabilize chemically reactive sludges that contain approximately 25 kilograms (55 pounds) of plutonium by heating to approximately 500 to 1000 degrees C (900 to 1800 degrees F). The resulting stable powder, mostly plutonium oxide with impurities, would be stored in the PFP vaults. The presence of chemically reactive plutonium-bearing sludges in the process gloveboxes poses a risk to workers from radiation exposure and limits the availability of storage space for future plant cleanup. Therefore, there is a need to stabilize the material into a form suitable for long-term storage. This proposed action would be an interim action, which would take place prior to completion of an Environmental Impact Statement for the PFP which would evaluate stabilization of all plutonium-bearing materials and cleanout of the facility. However, only 10 percent of the total quantity of plutonium in reactive materials is in the sludges, so this action will not limit the choice of reasonable alternatives or prejudice the Record of Decision of the Plutonium Finishing Plant Environmental Impact Statement

  14. Furnace System Testing to Support Lower-Temperature Stabilization of High Chloride Plutonium Oxide Items at the Hanford Plutonium Finishing Plant

    International Nuclear Information System (INIS)

    Schmidt, Andrew J.; Gerber, Mark A.; Fischer, Christopher M.; Elmore, Monte R.

    2003-01-01

    High chloride content plutonium (HCP) oxides are impure plutonium oxide scrap which contains NaCl, KCl, MgCl2 and/or CaCl2 salts at potentially high concentrations and must be stabilized at 950 C per the DOE Standard, DOE-STD-3013-2000. The chlorides pose challenges to stabilization because volatile chloride salts and decomposition products can corrode furnace heating elements and downstream ventilation components. Thermal stabilization of HCP items at 750 C (without water washing) is being investigated as an alternative method for meeting the intent of DOE STD 3013-2000. This report presents the results from a series of furnace tests conducted to develop material balance and system operability data for supporting the evaluation of lower-temperature thermal stabilization

  15. Plutonium metal burning facility

    International Nuclear Information System (INIS)

    Hausburg, D.E.; Leebl, R.G.

    1977-01-01

    A glove-box facility was designed to convert plutonium skull metal or unburned oxide to an oxide acceptable for plutonium recovery and purification. A discussion of the operation, safety aspects, and electrical schematics are included

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

    International Nuclear Information System (INIS)

    Pickering, S.

    1987-01-01

    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)

  17. Treatment of plutonium process residues by molten salt oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J. [Los Alamos National Lab., NM (United States); Heslop, M. [Naval Surface Warfare Center (United States). Indian Head Div.; Wernly, K. [Molten Salt Oxidation Corp. (United States)

    1999-04-01

    Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible {sup 238}Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na{sub 2}SO{sub 4}, Na{sub 3}PO{sub 4} and NaAsO{sub 2} or Na{sub 3}AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the {sup 238}Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox.

  18. Treatment of plutonium process residues by molten salt oxidation

    International Nuclear Information System (INIS)

    Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J.; Heslop, M.

    1999-01-01

    Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible 238 Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na 2 SO 4 , Na 3 PO 4 and NaAsO 2 or Na 3 AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the 238 Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox

  19. Aerosol generation by oxidation and combustion of plutonium and its compounds: literature survey

    International Nuclear Information System (INIS)

    Ballereau, P.

    1987-09-01

    Generation of aerosols by oxidation or combustion is one of the greatest risks due to plutonium. A review is made of the most interesting documents available on this topic. Following a brief study of plutonium oxydation conditions, characteristics of aerosols generated by accidents of fires involving metallic Pu and some of its compounds are assessed. Nuclear weapons are not included in this review [fr

  20. Plutonium estimation in the process solutions and oxide dissolved audit samplers by potentiometry using memo titrator

    International Nuclear Information System (INIS)

    Kumaraguru, K.; Shukla, Y.D.; Vijayan, K.; Ramamoorthy, N.; Jambunathan, U.; Kapoor, S.C.

    1990-01-01

    Potentiometric method is employed by using memotitrator coupled with combined electrode for the estimation of plutonium. The estimations are carried out on the process samples and the acid dissolved samples for auditing, in the concentration range of 5 g/l to 20 g/l. The chemical procedure is: i)oxidising plutonium to higher oxidation state by silver oxide, ii)reducing the same by adding excess ferrous, and iii)titrating potassium dichromate against the unreacted ferrous. The plutonium content is computed from ferrous consumed in the reaction. The average percentage error of the method is +/-0.27. The values obtained are in close agreement with those obtained by coulometry. (author)

  1. MIS High-Purity Plutonium Oxide Hydride Product 5501579 (SSR124): Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Veirs, Douglas Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stroud, Mary Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Berg, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Narlesky, Joshua Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Worl, Laura Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinex, Max A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Carillo, Alex [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2018-02-08

    A high-purity plutonium dioxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied with regard to gas generation and corrosion in a storage environment. Sample 5501579 represents process plutonium oxides from hydride oxide from Rocky Flats that are currently stored in 3013 containers. After calcination to 950°C, the material contained 87.42% plutonium with no major impurities. This study followed over time, the gas pressure of a sample with nominally 0.5 wt% water in a sealed container with an internal volume scaled to 1/500th of the volume of a 3013 container. Gas compositions were measured periodically over a six year period. The maximum observed gas pressure was 124 kPa. The increase over the initial pressure of 70 kPa was primarily due to generation of nitrogen and carbon dioxide gas. Hydrogen and oxygen were minor components of the headspace gas. At the completion of the study, the internal components of the sealed container showed signs of corrosion.

  2. Plutonium storage criteria

    Energy Technology Data Exchange (ETDEWEB)

    Chung, D. [Scientech, Inc., Germantown, MD (United States); Ascanio, X. [Dept. of Energy, Germantown, MD (United States)

    1996-05-01

    The Department of Energy has issued a technical standard for long-term (>50 years) storage and will soon issue a criteria document for interim (<20 years) storage of plutonium materials. The long-term technical standard, {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides,{close_quotes} addresses the requirements for storing metals and oxides with greater than 50 wt % plutonium. It calls for a standardized package that meets both off-site transportation requirements, as well as remote handling requirements from future storage facilities. The interim criteria document, {open_quotes}Criteria for Interim Safe Storage of Plutonium-Bearing Solid Materials{close_quotes}, addresses requirements for storing materials with less than 50 wt% plutonium. The interim criteria document assumes the materials will be stored on existing sites, and existing facilities and equipment will be used for repackaging to improve the margin of safety.

  3. Uranium plutonium oxide fuels

    International Nuclear Information System (INIS)

    Cox, C.M.; Leggett, R.D.; Weber, E.T.

    1981-01-01

    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 UO 2 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

  4. Electrochemical oxidation of nitrite on nanodiamond powder electrode

    Energy Technology Data Exchange (ETDEWEB)

    Chen, L.H.; Zang, J.B.; Wang, Y.H.; Bian, L.Y. [State Key Laboratory of Metastable Materials Science and Technology, Yanshan University, Qinhuangdao 066004 (China); College of Materials Science and Engineering, Yanshan University, Qinhuangdao 066004 (China)

    2008-03-10

    Nanodiamond (ND) powder electrodes were fabricated and the electrochemical properties were investigated in the solution containing nitrite in this article. This electrode exhibits substantial catalytic ability toward the oxidation of nitrite anions. The electrochemical oxidation mechanism of nitrite on the ND powder electrode is discussed. The oxidation of NaNO{sub 2} is a two-electron transfer process. The electrode reaction rate constant k is estimated to be 2.013 x 10{sup -4} cm/s and (1 - {alpha})n{sub {alpha}} is 0.1643. The peak current increases linearly with the rising of the concentration of NaNO{sub 2}. (author)

  5. Synthesis of vanadium oxide powders by evaporative decomposition of solutions

    International Nuclear Information System (INIS)

    Lawton, S.A.; Theby, E.A.

    1995-01-01

    Powders of the vanadium oxides V 2 O 4 , V 6 O 13 , and V 2 O 5 were produced by thermal decomposition of aqueous solutions of vanadyl sulfate hydrate in atmospheres of N 2 , H 2 mixed with N 2 , or air. The composition of the oxide powder was determined by the reactor temperature and gas composition. Residual sulfur concentrations in powders produced by decomposition at 740 C were less than 1 at.%, and these powders consisted of hollow, roughly spherical aggregates of particles less than 1 microm in diameter

  6. The effect of plutonium dioxide water surface coverage on the generation of hydrogen and oxygen

    Energy Technology Data Exchange (ETDEWEB)

    Veirs, Douglas K. [Los Alamos National Laboratory; Berg, John M. [Los Alamos National Laboratory; Crowder, Mark L. [Savannah River National Laboratory

    2012-06-20

    The conditions for the production of oxygen during radiolysis of water adsorbed onto plutonium dioxide powder are discussed. Studies in the literature investigating the radiolysis of water show that both oxygen and hydrogen can be generated from water adsorbed on high-purity plutonium dioxide powder. These studies indicate that there is a threshold in the amount of water below which oxygen is not generated. The threshold is associated with the number of monolayers of adsorbed water and is shown to occur at approximately two monolayers of molecularly adsorbed water. Material in equilibrium with 50% relative humidity (RH) will be at the threshold for oxygen generation. Using two monolayers of molecularly adsorbed water as the threshold for oxygen production, the total pressure under various conditions is calculated assuming stoichiometric production of hydrogen and oxygen. The specific surface area of the oxide has a strong effect on the final partial pressure. The specific surface areas resulting in the highest pressures within a 3013 container are evaluated. The potential for oxygen generation is mitigated by reduced relative humidity, and hence moisture adsorption, at the oxide surface which occurs if the oxide is warmer than the ambient air. The potential for oxygen generation approaches zero as the temperature difference between the ambient air and the material approaches 6 C.

  7. Plutonium in the marine environment

    International Nuclear Information System (INIS)

    Jarvis, N.V.; Linder, P.W.; Wade, P.W.

    1994-01-01

    The shipping of plutonium from Europe to Japan around the Cape is a contentious issue which has raised public concern that South Africa may be at risk to plutonium exposure should an accident occur. The paper describes the containers in which the plutonium (in the form of plutonium oxide, PuO 2 ) is housed and consequences of the unlikely event of these becoming ruptured. Wind-borne pollution is considered not to be a likely scenario, with the plutonium oxide particles more likely to remain practically insoluble and sediment. Plutonium aqueous and environmental chemistry is briefly discussed. Some computer modelling whereby plutonium oxide is brought into contact with seawater has been performed and the results are presented. The impact on marine organisms is discussed in terms of studies performed at marine dump sites and after the crash of a bomber carrying nuclear warheads in Thule, Greenland in 1968. Various pathways from the sea to land are considered in the light of studies done at Sellafield, a reprocessing plant in the United Kingdom. Some recent debates in the popular scientific press, such as that on the leukemia cluster at Sellafield, are described. Plutonium biochemistry and toxicity are discussed as well as medical histories of workers exposed to plutonium. 35 refs., 2 tabs., 1 fig

  8. Oxidized Metal Powders for Mechanical Shock and Crush Safety Enhancers; TOPICAL

    International Nuclear Information System (INIS)

    GARINO, TERRY J.

    2002-01-01

    The use of oxidized metal powders in mechanical shock or crush safety enhancers in nuclear weapons has been investigated. The functioning of these devices is based on the remarkable electrical behavior of compacts of certain oxidized metal powders when subjected to compressive stress. For example, the low voltage resistivity of a compact of oxidized tantalum powder was found to decrease by over six orders of magnitude during compaction between 1 MPa, where the thin, insulating oxide coatings on the particles are intact, to 10 MPa, where the oxide coatings have broken down along a chain of particles spanning the electrodes. In this work, the behavior of tantalum and aluminum powders was investigated. The low voltage resistivity during compaction of powders oxidized under various conditions was measured and compared. In addition, the resistivity at higher voltages and the dielectric breakdown strength during compaction were also measured. A key finding was that significant changes in the electrical properties persist after the removal of the stress so that a mechanical shock enhancer is feasible. This was verified by preliminary shock experiments. Finally, conceptual designs for both types of enhancers are presented

  9. Powder processing of high Tc oxide superconductors and their properties

    International Nuclear Information System (INIS)

    Vajpei, A.C.; Upadhyaya, G.S.

    1992-01-01

    Powder processing of ceramics is an established technology and in the area of high T c superconductors, its importance is felt even more significantly. The present monograph is an attempt in this direction to explore the perspectives and practice of powder processing routes towards control and optimization of the microstructure and pertinent properties of high T c oxide superconductors. The monograph consists of 6 chapters. After a very brief introduction (Chapter 1), Chapter 2 describes various classes of high T c oxide superconductors and their phase equilibria. Chapter 3 highlights the preparation of oxide superconductor powders through various routes and details their subtle distinctions. Chapter 4 briefly covers characterisation of the oxide superconductors, laying emphasis on the process-analysis and microstructure. Chapter 5 describes in detail various fabrication techniques for bulk superconductors through the powder routes. The last Chapter (Chapter 6) describing properties of bulk oxide superconductors, discusses the role of subtituents, compositional variations and processing methods on such properties. References are given at the end of each chapter. (orig.)

  10. 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.

  11. Plutonium-238 and plutonium-239 metabolism in dairy cows following ingestion of mixed oxides

    International Nuclear Information System (INIS)

    Patzer, R.G.; Mullen, A.A.; Sutton, W.W.; Potter, G.D.; Mosley, R.E.; Efurd, D.W.; Stalnaker, N.D.

    1985-01-01

    Dairy cows were given oral dosage of plutonium-238 and plutonium-239 dioxide particles in a study to determine the relative gastrointestinal absorption and tissue distribution of the nuclides. Two cows were given particles in which the two isotopes were homogeneously mixed within the particles. A third cow was given two batches of particles which contained either plutonium-238 or plutonium-239. Results indicate that, when the two isotopes of plutonium are homogeneous within the particles, there is no difference between plutonium-238 and plutonium-239 in the relative gastrointestinal absorption and tissue distribution

  12. Characterisation of oxidised aluminium powder: Validation of a new anodic oxidation bench

    Energy Technology Data Exchange (ETDEWEB)

    Gascoin, Nicolas, E-mail: Nicolas.Gascoin@univ-orleans.fr [PRISME Institute, Orleans University, 63 avenue de Lattre de Tassigny, 18020 Bourges (France); Gillard, Philippe; Baudry, Guillaume [PRISME Institute, Orleans University, 63 avenue de Lattre de Tassigny, 18020 Bourges (France)

    2009-11-15

    Aluminium powder is of major interest in many applications but it presents a risk due to its high explosibility, particularly when dispersed in air. The safety is directly linked to the particles oxidation because the Minimum Ignition Energy (MIE), which is required to initiate an Al dust explosion, increases with the oxide layer thickness. This study provides a controlled method to furnish reproducible homogeneous set of powder for such safety studies. Thanks to a new experimental bench, the influence on the oxidation rate of seven treatment parameters is investigated (current density, time of treatment, acid concentration, mass of powder, particles size, stirring, neutralisation by ammonia solution). The oxide content is plotted versus the current density, the time and the acid concentration to provide reference curves for further elaboration of oxidised powder. The particles size of sieved powder is measured before and after treatment by different methods (optical and Scanning Electron Microscopes, laser measurement). A high refinement of the powder in terms of size distribution is achieved thanks to the employed sieving. The present bench and the elaborated procedure are of great interest to provide well-calibrated oxidised powder directly available for safety studies. The time must be adjusted, depending on the wanted oxide content - from 2 to 18 wt.% - and the other treatment parameters must be kept constant: acid concentration (5 wt.%), current density (1 A dm{sup -2}), treated powder (20 g). In these conditions, the ratio of the oxide layer thickness on the particles diameter is found to be constant for a given oxide content whatever the particles size.

  13. Characterisation of oxidised aluminium powder: Validation of a new anodic oxidation bench

    International Nuclear Information System (INIS)

    Gascoin, Nicolas; Gillard, Philippe; Baudry, Guillaume

    2009-01-01

    Aluminium powder is of major interest in many applications but it presents a risk due to its high explosibility, particularly when dispersed in air. The safety is directly linked to the particles oxidation because the Minimum Ignition Energy (MIE), which is required to initiate an Al dust explosion, increases with the oxide layer thickness. This study provides a controlled method to furnish reproducible homogeneous set of powder for such safety studies. Thanks to a new experimental bench, the influence on the oxidation rate of seven treatment parameters is investigated (current density, time of treatment, acid concentration, mass of powder, particles size, stirring, neutralisation by ammonia solution). The oxide content is plotted versus the current density, the time and the acid concentration to provide reference curves for further elaboration of oxidised powder. The particles size of sieved powder is measured before and after treatment by different methods (optical and Scanning Electron Microscopes, laser measurement). A high refinement of the powder in terms of size distribution is achieved thanks to the employed sieving. The present bench and the elaborated procedure are of great interest to provide well-calibrated oxidised powder directly available for safety studies. The time must be adjusted, depending on the wanted oxide content - from 2 to 18 wt.% - and the other treatment parameters must be kept constant: acid concentration (5 wt.%), current density (1 A dm -2 ), treated powder (20 g). In these conditions, the ratio of the oxide layer thickness on the particles diameter is found to be constant for a given oxide content whatever the particles size.

  14. Water bath and air bath calorimeter qualification for measuring 3013 containers of plutonium oxide at the Hanford Plutonium Finishing Plant (PFP)

    International Nuclear Information System (INIS)

    WELSH, T.L.

    2003-01-01

    The purpose of this paper is to present qualification data generated from water and air-bath calorimeters measuring radioactive decay heat from plutonium oxide in DOE STD-3013-2000 (3013) containers at the Hanford Plutonium Finishing Plant (PFP). Published data concerning air and water bath calorimeters and especially 3013-qualified calorimeters is minimal at best. This paper will address the data from the measurement/qualification test plan, the heat standards used, and the calorimeter precision and accuracy results. The 3013 package is physically larger than earlier plutonium oxide storage containers, thereby necessitating a larger measurement chamber. To accommodate the measurements of the 3013 containers at PFP, Los Alamos National Laboratory (LANL) supplied a water bath dual-chambered unit and the Savannah River Technology Center (SRTC) provided two air-bath calorimeters. Both types of Calorimeters were installed in the analytical laboratory at PFP. The larger 3013 containers presented a new set of potential measurement problems: longer counting times, heat conductivity through a much larger container mass and wall thickness, and larger amounts of copper shot to assist sample thermal conductivity. These potential problems were addressed and included in the measurement/qualification test plan

  15. Conversion of plutonium-containing materials into borosilicate glass using the glass material oxidation and dissolution system

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

    1996-01-01

    The end of the cold war has resulted in excess plutonium-containing materials (PCMs) in multiple chemical forms. Major problems are associated with the long-term management of these materials: safeguards and nonproliferation issues; health, environment, and safety concerns; waste management requirements; and high storage costs. These issues can be addressed by conversion of the PCMs to glass: however, conventional glass processes require oxide-like feed materials. Conversion of PCMs to oxide-like materials followed by vitrification is a complex and expensive process. A new vitrification process has been invented, the Glass Material Oxidation and Dissolution System (GMODS) to allow direct conversion of PCMs to glass. GMODS directly converts metals, ceramics, and amorphous solids to glass; oxidizes organics with the residue converted to glass; and converts chlorides to borosilicate glass and a secondary sodium chloride stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium (a plutonium surrogate), Zircaloy, stainless steel, multiple oxides, and other materials to glass. Equipment options have been identified for processing rates between 1 and 100,000 t/y. Significant work, including a pilot plant, is required to develop GMODS for applications at an industrial scale

  16. 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...

  17. Plutonium bearing oxide fuels for recycling in thermal reactors and fast breeder reactors

    International Nuclear Information System (INIS)

    Cunningham, G.W.

    1977-01-01

    Programs carried out in the past two decades have established the technical feasibility of using plutonium as a fuel material in both water-cooled power reactors and sodium-cooled fast breeder reactors. The problem facing the technical community is basically one of demonstrating plutonium fuel recycle under strict conditions of public safety, accountability, personnel exposure, waste management, transportation and diversion or theft which are still evolving. In this paper only technical and economic aspects of high volume production and the demonstration program required are discussed. This paper discusses the role of mixed oxide fuels in light water reactors and the objectives of the LMFBR required for continual growth of nuclear power during the next century. The results of studies showing the impact of using plutonium on uranium requirements, power costs, and the market share of nuclear power are presented. The influence of doubling time and the introduction date of LMFBRs on the benefits to be derived by its commercial use are discussed. Advanced fuel development programs scoped to meet future commerical LMFBR fuel requirements are described. Programs designed to provide the basic technology required for using plutonium fuels in a manner which will satisfy all requirements for public acceptance are described. Included are the high exposure plutonium fabrication development program centered around the High Performance Fuels Laboratory being built at the Hanford Engineering Development Laboratory and the program to confirm the technology required for the production of mixed oxide fuels for light water reactors which is being coordinated by Savannah River Laboratories

  18. Elaboration and characterisation of yttrium oxide and hafnium oxide powders by the sol-gel process

    International Nuclear Information System (INIS)

    Hours, T.

    1988-01-01

    The two classical sol-gel processes, colloidal and polymeric are studied for the preparation of yttrium oxide and hafnium oxide high performance powders. In the colloidal process, controlled and reproducible conditions for the preparation of yttrium oxide and hafnium oxide sols from salts or alkoxides are developed and the hydrothermal synthesis monodisperse hafnium oxide colloids is studied. The polymeric process is studied with hafnium ethyl-hexylate, hydrolysis kinetics for controlled preparation of sols and gels is investigated. Each step of preparation is detailed and powders obtained are characterized [fr

  19. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO 2 and UO 2 ), typically containing 95% or more UO 2 . DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO 2 powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO 2 powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required

  20. Field characterization of plutonium aerosols in mixed-oxide fuel fabrication

    International Nuclear Information System (INIS)

    Newton, G.J.; Teague, S.V.; Yeh, H.C.

    1976-01-01

    Nuclear reactor fuel pellets of PuO 2 and UO 2 are fabricated within safety enclosures at Babcock and Wilcox's Parker Township Site near Apolla, Pa. Nineteen sample runs were taken from within glove boxes of aerosols formed during powder comminution and blending. Eight sampling runs were also taken of a centerless grinding operation during routine industrial operations. A small seven-stage cascade impactor and the Lovelace Aerosol Particle Separator (LAPS) were used to determine aerodynamic size distribution and gross alpha aerosol concentrations. The potential toxicity of inhaled plutonium originating in the nuclear fuel cycle following accidental releases of these aerosols and possible inhalation by industrial workers is considered

  1. A method of eliminating the surface defect in low-temperature oxidation powder added UO2 pellet

    International Nuclear Information System (INIS)

    Yoo, H. S.; Lee, S. J.; Kim, J. I.; Jeon, K. R.; Kim, J. W.

    2002-01-01

    A study on methods to eliminate surface defect shown in low-temperature oxidation powder added UO 2 pellet has been performed. Powders oxidized at 350 .deg. C for 4 hrs were prepared and mixed with UO 2 powder after crushing them. After being sintered, surfaces of the pellet were inspected both visually and optically. A large number of defects were observed on the surface of the specimens in which low-temperature oxidation powders were directly mixed or master mixed with UO 2 powder while both specimens produced from mixed powders including milled oxidation powders and powders that were milled totally after mixing had clean surfaces. However, optical examination showed considerably large defected pores in the milled oxidation powder added pellet and it was confirmed that the inner defects can be eliminated completely only when milling the entire mixture on UO 2 and low-temperature oxidation powder, but not by crushing only oxidation powder

  2. Method and apparatus for the production of metal oxide powder

    Science.gov (United States)

    Harris, Michael T.; Scott, Timothy C.; Byers, Charles H.

    1992-01-01

    The present invention provides a method for preparing metal oxide powder. A first solution, which is substantially organic, is prepared. A second solution, which is an aqueous solution substantially immiscible in the first solution, is prepared and delivered as drops to the first solution. The drops of the second solution are atomized by a pulsed electric field forming micro-drops of the second solution. Reagents in the first solution diffuse into and react with reactants in the micro-drops of the second solution forming metal hydroxide or oxalate particles. The metal hydroxide or metal oxalate particles are then recovered and dried to produce the metal oxide powder. An apparatus for preparing a metal oxide powder is also disclosed.

  3. Preparation of high purity plutonium oxide for radiochemistry instrument calibration standards and working standards

    International Nuclear Information System (INIS)

    Wong, A.S.; Stalnaker, N.D.

    1997-04-01

    Due to the lack of suitable high level National Institute of Standards and Technology (NIST) traceable plutonium solution standards from the NIST or commercial vendors, the CST-8 Radiochemistry team at Los Alamos National Laboratory (LANL) has prepared instrument calibration standards and working standards from a well-characterized plutonium oxide. All the aliquoting steps were performed gravimetrically. When a 241 Am standardized solution obtained from a commercial vendor was compared to these calibration solutions, the results agreed to within 0.04% for the total alpha activity. The aliquots of the plutonium standard solutions and dilutions were sealed in glass ampules for long term storage

  4. Plutonium inventories for stabilization and stabilized materials

    Energy Technology Data Exchange (ETDEWEB)

    Williams, A.K.

    1996-05-01

    The objective of the breakout session was to identify characteristics of materials containing plutonium, the need to stabilize these materials for storage, and plans to accomplish the stabilization activities. All current stabilization activities are driven by the Defense Nuclear Facilities Safety Board Recommendation 94-1 (May 26, 1994) and by the recently completed Plutonium ES&H Vulnerability Assessment (DOE-EH-0415). The Implementation Plan for accomplishing stabilization of plutonium-bearing residues in response to the Recommendation and the Assessment was published by DOE on February 28, 1995. This Implementation Plan (IP) commits to stabilizing problem materials within 3 years, and stabilizing all other materials within 8 years. The IP identifies approximately 20 metric tons of plutonium requiring stabilization and/or repackaging. A further breakdown shows this material to consist of 8.5 metric tons of plutonium metal and alloys, 5.5 metric tons of plutonium as oxide, and 6 metric tons of plutonium as residues. Stabilization of the metal and oxide categories containing greater than 50 weight percent plutonium is covered by DOE Standard {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides{close_quotes} December, 1994 (DOE-STD-3013-94). This standard establishes criteria for safe storage of stabilized plutonium metals and oxides for up to 50 years. Each of the DOE sites and contractors with large plutonium inventories has either started or is preparing to start stabilization activities to meet these criteria.

  5. The controlled-potential coulometric determination of plutonium based upon cerium oxidation and the Pu022+/Pu4+ valency change

    International Nuclear Information System (INIS)

    Phillips, G.; Crossley, D.; Venkataramana, P.

    1977-09-01

    Conditions have been established enabling plutonium to be oxidised quantitatively to the hexavalent state in the working compartment of a controlled-potential coulometric cell using electrogenerated ceric ion or excess ceric nitrate. The excess ceric ion is reduced in situ electrochemically without reduction of the hexavalent plutonium. The plutonium is then determined controlled-potential coulometrically by reduction to Pu 3+ followed by oxidation to Pu 4+ . The first oxidation step is conducted in molar nitric acid solution containing sulphamic acid but the coulometric determination step is conducted in molar sulphuric acid solution. The results obtained in the coulometric determination step were less satisfactory following oxidation with electrogenerated ceric ion rather than with chemically added ceric nitrate. Using the recommended conditions, 6 mg quantities of plutonium can be determined with an accuracy of 100.06% and a precision of 0.12% (coefficient of variation). The behaviour of chromium, manganese and vanadium impurity is reported. (author)

  6. Effect of microorganisms on the plutonium oxidation states

    International Nuclear Information System (INIS)

    Lukšienė, Benedikta; Druteikienė, Rūta; Pečiulytė, Dalia; Baltrūnas, Dalis; Remeikis, Vidmantas; Paškevičius, Algimantas

    2012-01-01

    Particular microbes from substrates at the low-level radioactive waste repository in the Ignalina NPP territory were exposed to 239 Pu (IV) at low pH under aerobic conditions. Pu(III) and Pu(IV) were separated and quantitatively evaluated using the modified anion exchange method and alpha spectrometry. Tested bacteria Bacillus mycoides and Serratia marcescens were more effective in Pu reduction than Rhodococcus fascians. Fungi Paecillomyces lilacinus and Absidia spinosa var. spinosa as well as bacterium Rhodococcus fascians did not alter the plutonium oxidation state. - Highlights: ► Particular microbes from low-level radioactive waste repository were exposed to Pu (IV). ► Some tested bacteria induced slight Pu (IV) reduction at low pH under aerobic conditions. ► Tested fungi did not show peculiarities to alter Pu oxidation state. ► The modified radiochemical method was applied to differentiate Pu oxidation states.

  7. Effect of ethylene oxide sterilization on the osteoinductivity of demineralized allograft bone powder

    International Nuclear Information System (INIS)

    Hamid Reza Aghayan; Babak Arjmand; Mehdi Golestani; Farokh Tirgari

    2008-01-01

    Full text: Ethylene oxide has been widely used for secondary sterilization of bone allograft to reduce the risk of infection and associated complications. In this study we investigate the effects of ethylene oxide gas sterilization on the osteoinductivity of demineralized bone Powder. Eighteen rats received two separate implants consisting of 30 mg aseptically prepared and 30 mg ethylene oxide-sterilized demineralized bone powder. The demineralized bone powder from each group was placed into two separate muscle pouch created in the paravertebral muscles of each rat. After 4 weeks each implantation site was removed with 0.5 cm normal tissue around the implant. Histological examination was done to determine the presence or absence of osteoinduction. All except one of eighteen aseptically prepared demineralized bone powder sites histologically contained new bone elements (94.4%) and fourteen (77.7%) of ethylene oxide sterilized demineralized bone powder sites showed evidence of new bone elements (p>0.05). There is no significant difference in osteoblast formation in two groups. We concluded that ethylene oxide sterilization in 42 degree C did not significantly reduce the osteoinductivity of demineralized bone powder. So ethylene oxide can be considered as a suitable but not perfect method for secondary sterilization of demineralized bone powder. (Author)

  8. Effect of cooling rate on achieving thermodynamic equilibrium in uranium–plutonium mixed oxides

    Energy Technology Data Exchange (ETDEWEB)

    Vauchy, Romain, E-mail: romain.vauchy@cea.fr [CEA, DEN, DTEC, Marcoule, 30207, Bagnols-sur-Cèze (France); CEA, DEN, DEC, Cadarache, 13108, Saint-Paul-lez-Durance (France); Belin, Renaud C.; Robisson, Anne-Charlotte [CEA, DEN, DEC, Cadarache, 13108, Saint-Paul-lez-Durance (France); Hodaj, Fiqiri [Univ. Grenoble Alpes, SIMAP, F-38000, Grenoble (France); CNRS, Grenoble INP, SIMAP, F-38000, Grenoble (France)

    2016-02-15

    In situ X-ray diffraction was used to study the structural changes occurring in uranium–plutonium mixed oxides U{sub 1−y}Pu{sub y}O{sub 2−x} with y = 0.15; 0.28 and 0.45 during cooling from 1773 K to room-temperature under He + 5% H{sub 2} atmosphere. We compare the fastest and slowest cooling rates allowed by our apparatus i.e. 2 K s{sup −1} and 0.005 K s{sup −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 U{sub 1−y}Pu{sub y}O{sub 2−x} uranium–plutonium mixed oxides.

  9. Formation of oxides particles in ferritic steel by using gas-atomized powder

    International Nuclear Information System (INIS)

    Liu Yong; Fang Jinghua; Liu Donghua; Lu Zhi; Liu Feng; Chen Shiqi; Liu, C.T.

    2010-01-01

    Oxides dispersion strengthened (ODS) ferritic steel was prepared by using gas-atomized pre-alloyed powder, without the conventional mechanical alloying process. By adjusting the volume content of O 2 in the gas atmosphere Ar, the O level in the ferritic powder can be well controlled. The O dissolves uniformly in the ferritic powder, and a very thin layer of oxides forms on the powder surface. After hot deformation, the primary particle boundaries, which retain after sintering, can be disintegrated and near fully dense materials can be obtained. The oxide layer on the powder surface has a significant effect on the microstructural evolution. It may prevent the diffusion in between the primary particles during sintering, and may dissolve and/or induce the nucleation of new oxides in the ferritic matrix during recrystallization. Two kinds of oxide particles are found in the ferritic steel: large (∼100 nm) Ti-rich and fine (10-20 nm) Y-Ti-rich oxides. The hardness of the ferritic steel increases with increasing annealing temperatures, however, decreases at 1400 deg. C, due to the coarsening of precipitates and the recrystallization microstructure.

  10. Pu(V) as the stable form of oxidized plutonium in natural waters

    International Nuclear Information System (INIS)

    Orlandini, K.A.; Penrose, W.R.; Nelson, D.M.

    1986-01-01

    This work presents analytical evidence supporting the proposition that Pu(V) is the sole or predominant form of oxidized plutonium in natural waters. Two independent methods, the selective adsorption of Pu(VI) by silica gel, and the somewhat less selective coprecipitation of Pu(V) with calcium carbonate, were developed to separate Pu(V) from Pu(VI). Measurements of ambient plutonium in several natural waters by these methods found only Pu(V). In laboratory tracer studies, Pu(VI) was shown to be highly unstable in dilute bicarbonate solution and in Lake Michigan water, reducing in first-order fashion to Pu(V). (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-22

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

  12. Method and apparatus for the production of metal oxide powder

    Science.gov (United States)

    Harris, M.T.; Scott, T.C.; Byers, C.H.

    1992-06-16

    The present invention provides a method for preparing metal oxide powder. A first solution, which is substantially organic, is prepared. A second solution, which is an aqueous solution substantially immiscible in the first solution, is prepared and delivered as drops to the first solution. The drops of the second solution are atomized by a pulsed electric field forming micro-drops of the second solution. Reagents in the first solution diffuse into and react with reactants in the micro-drops of the second solution forming metal hydroxide or oxalate particles. The metal hydroxide or metal oxalate particles are then recovered and dried to produce the metal oxide powder. An apparatus for preparing a metal oxide powder is also disclosed. 2 figs.

  13. Non-fertile fuels for burning weapons plutonium in thermal fission reactors

    International Nuclear Information System (INIS)

    Lombardi, C.; Mazzola, A.; Vettraino, F.

    1996-01-01

    In the last few years, the excess plutonium disposition has become ever more a topical and critical issue. As a matter of fact, more than 200 MT of plutonium coming from spent fuel reprocessing have been already stockpiled and over the next decade, under the already ratified agreements, another about 200 MT of weapon-grade plutonium are expected to be available from nuclear weapons dismantlement. On this basis, an ever growing plutonium production is no longer the goal and the already stored quantities should be burnt in power reactors by taking care that no new plutonium is generated under irradiation. This new outlook in considering plutonium has led many designers to reassess the Fast Breeder Reactors (FBR) role and shifting from breeder to burner machines perspective. Several solutions for burning plutonium have been so far proposed and discussed from the safeguards, proliferation resistance, environmental safety, technological background, economy and time schedule standpoint. A proposal for plutonium burning in commercial Pressurized Water Reactors (PWR) by using a non-fertile oxide-type fuel consisting of PuO 2 diluted in an inert matrix is reported hereafter. This solution appears to receive an ever growing interest in the nuclear community. In order not to produce new plutonium during irradiation an innovative U-free fuel is being researched, based on an inert matrix which will consist in a mixed compound of inert oxides, such as ZrO 2 , Al2O 3 , MgO, CeO 2 where the plutonium oxide is dispersed in. The matrix will fulfill the following requirements: good chemical compatibility, acceptable thermal conductivity, good nuclear properties, good stability under irradiation, good dissolution resistance. The plutonium relative content will be comparable to that used in MOX fuel. The fuel is expected to be characterized by a high chemical stability (rock-like fuel), so that after discharge from reactor and adequate cooling time, it can be considered a High Level

  14. A simple and rapid method of purification of impure plutonium oxide

    International Nuclear Information System (INIS)

    Michael, K.M.; Rakshe, P.R.; Dharmpurikar, G.R.; Thite, B.S.; Lokhande, Manisha; Sinalkar, Nitin; Dakshinamoorthy, A.; Munshi, S.K.; Dey, P.K.

    2007-01-01

    Impure plutonium oxides are conventionally purified by dissolution in HNO 3 in presence of HF followed by ion exchange separation and oxalate precipitation. The method is tedious and use of HF enhances corrosion of the plant equipment's. A simple and rapid method has been developed for the purification of the oxide by leaching with various reagents like DM water, NaOH and oxalic acid. A combination of DM water followed by hot leaching with 0.4 M oxalic acid could bring down the impurity levels in the oxide to the desired level required for fuel fabrication. (author)

  15. Oxide Dispersion Strengthened Iron Aluminide by CVD Coated Powders

    Energy Technology Data Exchange (ETDEWEB)

    Asit Biswas Andrew J. Sherman

    2006-09-25

    This I &I Category2 program developed chemical vapor deposition (CVD) of iron, aluminum and aluminum oxide coated iron powders and the availability of high temperature oxidation, corrosion and erosion resistant coating for future power generation equipment and can be used for retrofitting existing fossil-fired power plant equipment. This coating will provide enhanced life and performance of Coal-Fired Boilers components such as fire side corrosion on the outer diameter (OD) of the water wall and superheater tubing as well as on the inner diameter (ID) and OD of larger diameter headers. The program also developed a manufacturing route for readily available thermal spray powders for iron aluminide coating and fabrication of net shape component by powder metallurgy route using this CVD coated powders. This coating can also be applid on jet engine compressor blade and housing, industrial heat treating furnace fixtures, magnetic electronic parts, heating element, piping and tubing for fossil energy application and automotive application, chemical processing equipment , heat exchanger, and structural member of aircraft. The program also resulted in developing a new fabrication route of thermal spray coating and oxide dispersion strengthened (ODS) iron aluminide composites enabling more precise control over material microstructures.

  16. White paper on possible inclusion of mixed plutonium-uranium oxides in DOE-STD-3013-96

    International Nuclear Information System (INIS)

    Haschke, J.M.; Venetz, T.; Szempruch, R.; McClard, J.W.

    1997-11-01

    This report assesses stabilization issues concerning mixed plutonium-uranium oxides containing 50 mass % Pu. Possible consequences of uranium substitution on thermal stabilization, specific surface areas, moisture readsorption behavior, loss-on-ignition analysis, and criticality safety of the oxide are examined and discussed

  17. Plutonium oxide shipment report

    International Nuclear Information System (INIS)

    1979-01-01

    While following procedures for unloading shipping containers containing plutonium oxide, SRP personnel experienced problems. While using a pipe cutter to cut through the wall of the inner container, the pipe section fell to the floor. Three empty food cans in the bottom of the inner canister also fell to the floor and a puff of smoke was observed. Personnel were evacuated and contamination was detected in the room. As a result of the investigations conducted by Westinghouse and SRP, thermal effects, food can coatings, and fuel volatiles were eliminated as the cause of the problem. Helium used to leak test the RL070 shipping container seals entered the inner canister through two weld defects resulting in a pressurization of the contents. When the end cap was removed, the inner canister vented rapidly, the food cans did not, thus creating a differential pressure across the food cans. This caused the food cans to swell. It was recommended that a dye penetrant test of all inner container welds be added. Additional unloading procedures were also recommended

  18. Inhaled plutonium oxide in dogs

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    This project is concerned with long-term experiments to determine the life-span dose-effect relationships of inhaled 239 PuO 2 and 238 PuO 2 in beagles. The data will be used to estimate the health effects of inhaled transuranics. The tissue distribution of plutonium, radiation effects in the lung and hematologic changes in plutonium-exposed beagles with lung tumors were evaluated

  19. Comparison of NDA and DA measurement techniques for excess plutonium powders at the Hanford Site: Statistical design and heterogeneity testing

    International Nuclear Information System (INIS)

    Welsh, T.L.; McRae, L.P.; Delegard, C.H.; Liebetrau, A.M.; Johnson, W.C.; Theis, W.; Lemaire, R.J.; Xiao, J.

    1995-06-01

    Quantitative physical measurements are a n component of the International Atomic Energy Agency (IAEA) nuclear material m ampersand guards verification regime. In December 1994, LA.FA safeguards were initiated on an inventory of excess plutonium powder items at the Plutonium Finishing Plant, Vault 3, on the US Department of Energy's Hanford Site. The material originl from the US nuclear weapons complex. The diversity of the chemical form and the heterogenous physical form of this inventory were anticipated to challenge the precision and accuracy of quantitative destructive analytical techniques. A sampling design was used to estimate the degree of heterogeneity of the plutonium content of a variety of inventory items. Plutonium concentration, the item net weight, and the 240 Pu content were among the variables considered in the design. Samples were obtained from randomly selected location within each item. Each sample was divided into aliquots and analyzed chemically. Operator measurements by calorimetry and IAEA measurements by coincident neutron nondestructive analysis also were performed for the initial physical inventory verification materials and similar items not yet under IAEA safeguards. The heterogeneity testing has confirmed that part of the material is indeed significantly heterogeneous; this means that precautionary measures must be taken to obtain representative samples for destructive analysis. In addition, the sampling variability due to material heterogeneity was found to be comparable with, or greater than, the variability of the operator's calorimetric measurements

  20. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  1. Plutonium uptake by the green alga Scenedesmus obliquus (Turp) Kutz, as a function of isotope and oxidation state

    Energy Technology Data Exchange (ETDEWEB)

    Tkacik, M.F.

    1977-01-01

    This study was designed to determine the effect of plutonium chemical valence state on the availability of small concentrations of /sup 238/Pu and /sup 239/Pu to algae. The uptake experiments involved the green alga Scenedesmus obliquus, grown in batch cultures. Plutonium concentrations accumulated by this alga were linearly related to plutonium concentrations. There was no significant difference (rho = 0.05) in algal plutonium accumulations, on a mass basis, of either /sup 238/Pu or /sup 239/Pu in either Pu/sup +4/ or Pu/sup +6/ oxidation state at the concentrations studied.

  2. The inhalation of insoluble iron oxide particles in the sub-micron ranges. Part II - Plutonium-237 labelled aerosols

    International Nuclear Information System (INIS)

    Waite, D.A.; Ramsden, D.

    1971-10-01

    The results of a series of inhalation studies using iron oxide particles in the size range 0.1 to 0.3 um (count median diameter) are described. In this series the aerosols were labelled with plutonium 237. In vivo detection, excretion analysis and crude location studies were obtainable and the results compared to the earlier studies using chromium 51 labelled aerosols. Plutonium 237 can be considered as a simulator for plutonium 239 and attempts are made to extrapolate the results to the problem of the estimation of plutonium 239 in the human lung. (author)

  3. Plutonium uptake by the green alga Scenedesmus obliquus (Turp) Kutz, as a function of isotope and oxidation state

    International Nuclear Information System (INIS)

    Tkacik, M.F.

    1977-01-01

    This study was designed to determine the effect of plutonium chemical valence state on the availability of small concentrations of 238 Pu and 239 Pu to algae. The uptake experiments involved the green alga Scenedesmus obliquus, grown in batch cultures. Plutonium concentrations accumulated by this alga were linearly related to plutonium concentrations. There was no significant difference (rho = 0.05) in algal plutonium accumulations, on a mass basis, of either 238 Pu or 239 Pu in either Pu +4 or Pu +6 oxidation state at the concentrations studied

  4. Effect of microorganisms on the plutonium oxidation states

    Energy Technology Data Exchange (ETDEWEB)

    Luksiene, Benedikta, E-mail: bena@ar.fi.lt [Center for Physical Sciences and Technology, Savanoriu ave 231, LT-02300 Vilnius (Lithuania); Druteikiene, Ruta [Center for Physical Sciences and Technology, Savanoriu ave 231, LT-02300 Vilnius (Lithuania); Peciulyte, Dalia [Nature Research Centre, Akademijos street 2, LT-08412 Vilnius (Lithuania); Baltrunas, Dalis; Remeikis, Vidmantas [Center for Physical Sciences and Technology, Savanoriu ave 231, LT-02300 Vilnius (Lithuania); Paskevicius, Algimantas [Nature Research Centre, Akademijos street 2, LT-08412 Vilnius (Lithuania)

    2012-03-15

    Particular microbes from substrates at the low-level radioactive waste repository in the Ignalina NPP territory were exposed to {sup 239}Pu (IV) at low pH under aerobic conditions. Pu(III) and Pu(IV) were separated and quantitatively evaluated using the modified anion exchange method and alpha spectrometry. Tested bacteria Bacillus mycoides and Serratia marcescens were more effective in Pu reduction than Rhodococcus fascians. Fungi Paecillomyces lilacinus and Absidia spinosa var. spinosa as well as bacterium Rhodococcus fascians did not alter the plutonium oxidation state. - Highlights: Black-Right-Pointing-Pointer Particular microbes from low-level radioactive waste repository were exposed to Pu (IV). Black-Right-Pointing-Pointer Some tested bacteria induced slight Pu (IV) reduction at low pH under aerobic conditions. Black-Right-Pointing-Pointer Tested fungi did not show peculiarities to alter Pu oxidation state. Black-Right-Pointing-Pointer The modified radiochemical method was applied to differentiate Pu oxidation states.

  5. Treatment of plutonium contaminations

    International Nuclear Information System (INIS)

    Lafuma, J.

    1983-01-01

    Three kinds of plutonium contaminations were considered: skin contamination; contaminated wounds; contamination by inhalation. The treatment of these contaminations was studied for insoluble (oxide and metal forms) and soluble plutonium (complexes). The use of DTPA and therapeutic problems encountered with stable plutonium complexes were analyzed. The new possibilities of internal decontamination using Puchel and LICAM were evaluated [fr

  6. Graphene oxide powders with different oxidation degree, prepared by synthesis variations of the Hummers method

    International Nuclear Information System (INIS)

    Guerrero-Contreras, Jesus; Caballero-Briones, F.

    2015-01-01

    Graphene oxide (GO) powders with different oxidation degree estimated through the relative intensity of the infrared absorption bands related to oxygen containing groups were prepared through variations of the Hummers method. The GO powders were analyzed by Transmission Electron Microscopy, Energy dispersive spectroscopy, X-ray Photoelectron Spectroscopy, Fourier Transform Infrared Spectroscopy, Raman spectroscopy, X-ray Diffraction, UV–VIS spectroscopy and Electrical Resistance measurements. Several square micron GO sheets with low wrinkling were obtained. Oxygen to carbon ratio is around 0.2 in all the samples although a strong variance in the relative intensity of the oxygen related infrared bands is evident. Thus, the oxidation degree was estimated from the FTIR measurements using the quotient between the C–O related bands area to the total area under the spectra. FTIR shows presence of hydroxyl (–OH), epoxy (C–O–C), carboxyl (–COOH) and carbonyl (C=O) moieties and evidence of intermolecular interactions between adjacent groups. These interactions influence the exfoliation degree, the absorbance of the GO suspensions, as well as the electrical resistance, while the crystalline domain sizes, estimated from XRD and Raman do not show a noticeable behavior related with the composition and molecular structure. The results indicate that the electrical resistance is influenced mainly by the surface chemistry of the GO powders and not only by the O/C ratio. The control of the surface chemistry of GO powders would allow their use as additives in organic bulk heterojunction solar cells with enhanced photoconversion efficiency. - Highlights: • Powders of graphene oxide with different oxidation degree were prepared through variations of the Hummers method. • Raman spectroscopy and XRD demonstrated similar crystallite domain size in the samples. • Electrical resistance, exfoliation degree and optical absorption depend on the molecular structure.

  7. Graphene oxide powders with different oxidation degree, prepared by synthesis variations of the Hummers method

    Energy Technology Data Exchange (ETDEWEB)

    Guerrero-Contreras, Jesus; Caballero-Briones, F., E-mail: fcaballero@ipn.mx

    2015-03-01

    Graphene oxide (GO) powders with different oxidation degree estimated through the relative intensity of the infrared absorption bands related to oxygen containing groups were prepared through variations of the Hummers method. The GO powders were analyzed by Transmission Electron Microscopy, Energy dispersive spectroscopy, X-ray Photoelectron Spectroscopy, Fourier Transform Infrared Spectroscopy, Raman spectroscopy, X-ray Diffraction, UV–VIS spectroscopy and Electrical Resistance measurements. Several square micron GO sheets with low wrinkling were obtained. Oxygen to carbon ratio is around 0.2 in all the samples although a strong variance in the relative intensity of the oxygen related infrared bands is evident. Thus, the oxidation degree was estimated from the FTIR measurements using the quotient between the C–O related bands area to the total area under the spectra. FTIR shows presence of hydroxyl (–OH), epoxy (C–O–C), carboxyl (–COOH) and carbonyl (C=O) moieties and evidence of intermolecular interactions between adjacent groups. These interactions influence the exfoliation degree, the absorbance of the GO suspensions, as well as the electrical resistance, while the crystalline domain sizes, estimated from XRD and Raman do not show a noticeable behavior related with the composition and molecular structure. The results indicate that the electrical resistance is influenced mainly by the surface chemistry of the GO powders and not only by the O/C ratio. The control of the surface chemistry of GO powders would allow their use as additives in organic bulk heterojunction solar cells with enhanced photoconversion efficiency. - Highlights: • Powders of graphene oxide with different oxidation degree were prepared through variations of the Hummers method. • Raman spectroscopy and XRD demonstrated similar crystallite domain size in the samples. • Electrical resistance, exfoliation degree and optical absorption depend on the molecular structure.

  8. Preparation and study of the critical-mass-free plutonium ceramics with neutron poisons Hf, Gd and Li

    International Nuclear Information System (INIS)

    Timoefeeva, L.F.; Orlov, V.K.; Malyukov, E.E.; Molomin, V.I.; Zhmak, V.A.; Semova, E.A.; Shishkov, N.V.; Nadykto, B.A.

    2002-01-01

    Powder sintering was used to produce homogeneous type oxide ceramics of Pu with Hf, Gd and Li 6 . In all the ceramics, there is the number of neutron poison (Hf, Gd and Li) atoms per plutonium atom needed, according to the physical calculation, for them to be free of critical mass. PuO 2 stabilizers high-temperature modifications of cubic HfO 2 or hexagonal Gd 2 O 3 , however, at the ratio given by the physical calculation, the plutonium is insufficient for their full stabilization. Addition of yttrium oxide as an additive stabilizing the fcc phase of HfO 2 resulted in cubic solid solution (Pu, Hf, Y)O 2-x . Pu/Li/Hf and Pu/Li/Si ceramics produced by sintering of PuO 2 and compound Li 2 HfO 3 or 6 Li 4 SiO 4 powders is characterized with presence of two phases. The method of differential thermal analysis demonstrated the phase stability of (Pu-Hf, Pu-Gd, Pu-Li-Hf) oxide ceramics in the 20-1500degC temperature range. Ceramic (Pu/Li/Si) has several endothermal effects. Tests in boiling water solutions of various composition suggest that the specimens of Pu, Hf oxides and ternary oxides (Pu, Hf, Y)O 2 are less stable in weakly acidic media than in weakly alkaline medium and distilled water. The obtained results were used as a basis to estimate the assumed solid solution region boundaries for binary Hf, Pu and ternary Hf, Pu, Y oxides on the side of HfO 2 . (author)

  9. Baseline process description for simulating plutonium oxide production for precalc project

    Energy Technology Data Exchange (ETDEWEB)

    Pike, J. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-10-26

    Savannah River National Laboratory (SRNL) started a multi-year project, the PreCalc Project, to develop a computational simulation of a plutonium oxide (PuO2) production facility with the objective to study the fundamental relationships between morphological and physicochemical properties. This report provides a detailed baseline process description to be used by SRNL personnel and collaborators to facilitate the initial design and construction of the simulation. The PreCalc Project team selected the HB-Line Plutonium Finishing Facility as the basis for a nominal baseline process since the facility is operational and significant model validation data can be obtained. The process boundary as well as process and facility design details necessary for multi-scale, multi-physics models are provided.

  10. Plutonium separation by reduction stripping. Application to processing of mixed oxide (U,Pu)O2 fuel fabrication wastes

    International Nuclear Information System (INIS)

    Arnal, Thierry; Cousinou, Gerard; Ganivet, Michel.

    1978-11-01

    A procedure is described for separating plutonium from a uranium VI and plutonium IV mixture contained in an organic phase (tributyl phosphate diluted in dodecane). This separation is obtained by extracting the plutonium III using two organic reducers: hydrazine and paraminophenol. Paraminophenol has excellent reducing qualities, similar to those of ferrous sulphamate, but has the added advantage of not contaminating extracted plutonium. This procedure is currently used in processing production wastes from mixed oxide (U,Pu)O 2 fuels; the installation using this procedure is described in detail in this paper. Operating results show the remarkable efficiency of this procedure: the separated plutonium and uranium mass flows have been increased to 185 and 350 g.h -1 respectively; the uranium contains less than 0.1 ppm of plutonium on completion of the purification cycle [fr

  11. 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.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

    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

  13. Examination of Zinc Oxide Nanoparticles as a Fluorescent Fingerprint Detection Powder

    International Nuclear Information System (INIS)

    Tun Tun Lin

    2010-12-01

    Detection of latent fingerprint was performed using zinc oxide nanoparticles which were produced by simple and efficient method in aqueous media from zinc nitrate. Synthesized ZnO nanoparticles were characterized by XRD, SEM and AFM for ZnO purification and particle size examination. In this paper an effort has been made to compare the results of using ZnO nanoparticles and conventional fingerprint powders such as ZnO bulk powder, CaO, TiO2, printer toner powder and graphite. Fingerprints on different materials were also examined by the use of ZnO and Graphite powder, which is currently used in the Central Intelligence Department of Myanmar Police Force.From this research, it was observed that zinc oxide nanoparticles powder produced a much clearer picture of the fingerprints, compared to conventional powders and it has very good quality at sticking to the fingerprint residue but not to the background surface.

  14. Solvent anode for plutonium purification

    International Nuclear Information System (INIS)

    Bowersox, D.F.; Fife, K.W.; Christensen, D.C.

    1986-01-01

    The purpose of this study is to develop a technique to allow complete oxidation of plutonium from the anode during plutonium electrorefining. This will eliminate the generation of a ''spent'' anode heel which requires further treatment for recovery. Our approach is to employ a solvent metal in the anode to provide a liquid anode pool throughout electrorefining. We use molten salts and metals in ceramic crucibles at 700 0 C. Our goal is to produce plutonium metal at 99.9% purity with oxidation and transfer of more than 98% of the impure plutonium feed metal from the anode into the salt and product phases. We have met these criteria in experiments on the 100 to 1000 g scale. We plan to scale our operations to 4 kg of feed plutonium and to optimize the process parameters

  15. The 871 keV gamma ray from 17O and the identification of plutonium oxide

    International Nuclear Information System (INIS)

    Peurrung, Anthony; Arthur, Richard; Elovich, Robert; Geelhood, Bruce; Kouzes, Richard; Pratt, Sharon; Scheele, Randy; Sell, Richard

    2001-01-01

    Disarmament agreements and discussions between the United States and the Russian Federation for reducing the number of stockpiled nuclear weapons require verification of the origin of materials as having come from disassembled weapons. This has resulted in the identification of measurable 'attributes' that characterize such materials. It has been proposed that the 871 keV gamma ray of 17 O can be observed as an indicator of the unexpected presence of plutonium oxide, as opposed to plutonium metal, in such materials. We have shown that the observation of the 871 keV gamma ray is not a specific indicator of the presence of the oxide, but rather indicates the presence of nitrogen

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

    International Nuclear Information System (INIS)

    1967-01-01

    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 PuO 2 -MeO x (Me=Be, Mg, Al, Si, W, Th, Eu, Zr, Ce) systems. 167 refs, 27 figs, 17 tabs

  17. Measurements of the oxidation state and concentration of plutonium in interstitial waters of the Irish Sea

    International Nuclear Information System (INIS)

    Nelson, D.M.; Lovett, M.B.

    1980-01-01

    The question of plutonium movement in interstitial waters resulting from diffusion along concentration gradients or from advective flow is addressed. The results of measurements of both the concentration and the oxidation state of plutonium in interstitial water collected from sediments near the Windscale discharge, in the solid phases of these sediments and in seawater and suspended solids collected at the coring locations are discussed

  18. Modified titrimetric determination of plutonium using photometric end-point detection

    International Nuclear Information System (INIS)

    Baughman, W.J.; Dahlby, J.W.

    1980-04-01

    A method used at LASL for the accurate and precise assay of plutonium metal was modified for the measurement of plutonium in plutonium oxides, nitrate solutions, and in other samples containing large quantities of plutonium in oxidized states higher than +3. In this modified method, the plutonium oxide or other sample is dissolved using the sealed-reflux dissolution method or other appropriate methods. Weighed aliquots, containing approximately 100 mg of plutonium, of the dissolved sample or plutonium nitrate solution are fumed to dryness with an HC1O 4 -H 2 SO 4 mixture. The dried residue is dissolved in dilute H 2 SO 4 , and the plutonium is reduced to plutonium (III) with zinc metal. The excess zinc metal is dissolved with HCl, and the solution is passed through a lead reductor column to ensure complete reduction of the plutonium to plutonium (III). The solution, with added ferroin indicator, is then titrated immediately with standardized ceric solution to a photometric end point. For the analysis of plutonium metal solutions, plutonium oxides, and nitrate solutions, the relative standard deviation are 0.06, 0.08, and 0.14%, respectively. Of the elements most likely to be found with the plutonium, only iron, neptunium, and uranium interfere. Small amounts of uranium and iron, which titrate quantitatively in the method, are determined by separate analytical methods, and suitable corrections are applied to the plutonium value. 4 tables, 4 figures

  19. 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

    2016-03-07

    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.

  20. Metallization of uranium oxide powders by lithium reduction

    International Nuclear Information System (INIS)

    Kim, I. S.; Seo, J. S.; Oh, S. C.; Hong, S. S.; Lee, W. K.

    2002-01-01

    Laboratory scale experiments on the reduction of uranium oxide powders into metal by lithium were performed in order to determine the equipment setup and optimum operation conditions. The method of filtration using the porous magnesia filter was introduced to recover uranium metal powders produced. Based on the laboratory scale experimental results, mock-up scale (20 kg U/batch) metallizer was designed and made. The applicability to the metallization process was estimated with respect to the thermal stability of the porous magnesia filter in the high temperature molten salt, the filtration of the fine uranium metal powders, and the operability of the equipment

  1. Plutonium gastrointestinal absorption by adults baboons

    International Nuclear Information System (INIS)

    Lataillade, G.; Madic, C.; Willemot, J.M.; Germain, P.; Colle, C.; Metivier, H.

    1991-01-01

    Gastrointestinal absorption of plutonium was investigated in baboons after ingestion of plutonium solution (oxidation states III; IV; V; VI), and plutonium incorporated in soya bean and winkles. We studied the effects of oxidation state and ingested mass for masses ranging from 0.35 μg to 51.6 x 10 +3 μg (4 x 10 -2 to 7776 μg of plutonium per kg of body weight). No clear increase in plutonium retention was shown for concentrations of plutonium smaller than 1 μg/kg. From 1 μg/kg to 1 mg/kg no effects of mass or oxidation state was observed and the mean fractional retention value was 10 -4 . For ingested masses higher than 1 mg/kg the fractional retention values respectively increased for Pu(V) and Pu(III) to (0.9 + 0.2) x 10 -2 and (7.4 + 4.1) x 10 -4 of the ingested mass. This increase might be due to the weak hydrolysis of these oxidation states which would increase gastrointestinal absorption by decrease of hydroxide formation. The fraction of plutonium retained after ingestion of soya bean was (3.0 + 0.5) x 10 -4 about 3 fold higher than the value for 238 Pu nitrate solution. No clear increase in plutonium retention was shown after ingestion of winkles containing 238 Pu. In conclusion, except for high masses of ingested Pu, the retention of which could reach 1% of the ingested dose, our results show that the gastrointestinal transfer factor of 10 -4 proposed by ICRP for gastrointestinal absorption of soluble form of Pu is acceptable, but 10 -3 would provide better safety margin [fr

  2. Direct oxide reduction (DOR) solvent salt recycle in pyrochemical plutonium recovery operations

    International Nuclear Information System (INIS)

    Fife, K.W.; Bowersox, D.F.; Davis, C.C.; McCormick, E.D.

    1987-02-01

    One method used at Los Alamos for producing plutonium metal is to reduce the oxide with calcium metal in molten CaCl 2 at 850 0 C. The solvent CaCl 2 from this reduction step is currently discarded as low-level radioactive waste because it is saturated with the reaction by-product, CaO. We have developed and demonstrated a molten salt technique for rechlorinating the CaO, thereby regenerating the CaCl 2 and incorporating solvent recycle into the batch PuO 2 reduction process. We discuss results from the process development experiments and present our plans for incorporating the technique into an advanced design for semicontinuous plutonium metal production

  3. Disposition of Uranium -233 (sup 233U) in Plutonium Metal and Oxide at the Rocky Flats Environmental Technology Site

    International Nuclear Information System (INIS)

    Freiboth, Cameron J.; Gibbs, Frank E.

    2000-01-01

    This report documents the position that the concentration of Uranium-233 ( 233 U) in plutonium metal and oxide currently stored at the DOE Rocky Flats Environmental Technology Site (RFETS) is well below the maximum permissible stabilization, packaging, shipping and storage limits. The 233 U stabilization, packaging and storage limit is 0.5 weight percent (wt%), which is also the shipping limit maximum. These two plutonium products (metal and oxide) are scheduled for processing through the Building 371 Plutonium Stabilization and Packaging System (PuSPS). This justification is supported by written technical reports, personnel interviews, and nuclear material inventories, as compiled in the ''History of Uranium-233 ( 233 U) Processing at the Rocky Flats Plant In Support of the RFETS Acceptable Knowledge Program'' RS-090-056, April 1, 1999. Relevant data from this report is summarized for application to the PuSPS metal and oxide processing campaigns

  4. Design of the Laboratory-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Lumetta, Gregg J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Meier, David E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tingey, Joel M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Delegard, Calvin H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Edwards, Matthew K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Orton, Robert D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rapko, Brian M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smart, John E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-05-01

    This report describes a design for a laboratory-scale capability to produce plutonium oxide (PuO2) for use in identifying and validating nuclear forensics signatures associated with plutonium production, as well as for use as exercise and reference materials. This capability will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including PuO2 dissolution, purification of the Pu by ion exchange, precipitation, and re-conversion to PuO2 by calcination.

  5. Direct reduction of plutonium from dicesium hexachloroplutonate

    International Nuclear Information System (INIS)

    Averill, W.A.; Boyd, T.E.

    1991-01-01

    The Rocky Flats Plant produces dicesium hexachloroplutonate (DCHP) primarily as a reagent in the molten salt extraction of americium from plutonium metal. DCHP is precipitated from aqueous chloride solutions derived from the leaching of process residues with a high degree of selectivity. DCHP is a chloride salt of plutonium, while the traditional aqueous precipitate is a hydrated oxide. Plutonium metal preparation from the oxide involves either the conversion of oxide to a halide followed by metallothermic reduction or direct reduction of the oxide using a flux. Either method generates at least three times as much radioactively contaminated waste as metal produced. Plutonium concentration by DCHP precipitation, however, produces a chloride salt that can be reduced using calcium metal at a temperature of approximately 1000K. In this paper the advantages and limitations of this process are discussed

  6. Analytical determination of plutonium in the presence of complexing agents

    International Nuclear Information System (INIS)

    Nebel', D.

    1975-01-01

    The complexing of Pu(4) and Pu(6) with citrate and lactate were studied to establish possibilities which are provided by the complexing for analytical determination of plutonium in a solution. Plutonium in the sample analyzed is preliminarily reduced up to trivalent state. Then trivalent plutonium is oxidized up to tetravalent on the platinum electrode at +1.05 V. To avoid further oxidation complexing agents are added. The value of pH=2-4 is maintained to stabilize the oxidation degree. If hexavalent plutonium should be obtained, the oxidation is performed at the voltage of +1.5 V in the absence of complexing agents. Based on a detailed study of absorption spectra of plutonium complexes of various oxidation degrees it is stated that the most suitable for analytical purposes is the oxidation degree of Pu=4

  7. Managing plutonium in Britain. Current options[Mixed oxide nuclear fuels; Nuclear weapons

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This is the report of a two day meeting to discuss issues arising from the reprocessing of plutonium and production of mixed oxide nuclear fuels in Britain. It was held at Charney Manor, near Oxford, on June 25 and 26, 1998, and was attended by 35 participants, including government officials, scientists, policy analysts, representatives of interested NGO's, journalists, a Member of Parliament, and visiting representatives from the US and Irish governments. The topic of managing plutonium has been a consistent thread within ORG's work, and was the subject of one of our previous reports, CDR 12. This particular seminar arose out of discussions earlier in the year between Dr. Frank Barnaby and the Rt. Hon. Michael Meacher MP, Minister for the Environment. With important decisions about the management of plutonium in Britain pending, ORG undertook to hold a seminar at which all aspects of the subject could be aired. A number of on-going events formed the background to this initiative. The first was British Nuclear Fuels' [BNFL] application to the Environment Agency to commission a mixed oxide fuel [MOX] plant at Sellafield. The second was BNFL's application to vary radioactive discharge limits at Sellafield. Thirdly, a House of Lords Select Committee was in process of taking evidence, on the disposal of radioactive waste. Fourthly, the Royal Society, in a recent report entitled Management of Separated Plutonium, recommended that 'the Government should commission a comprehensive review... of the options for the management of plutonium'. Four formal presentations were made to the meeting, on the subjects of Britain's plutonium policy, commercial prospects for plutonium use, problems of plutonium accountancy, and the danger of nuclear terrorism, by experts from outside the nuclear industry. It was hoped that the industry's viewpoint would also be heard, and BNFL were invited to present a paper, but declined on the grounds that they

  8. Properties of plutonium

    International Nuclear Information System (INIS)

    Ahn, Jin Su; Yoon, Hwan Ki; Min, Kyung Sik; Kim, Hyun Tae; Ahn, Jong Sung; Kwag, Eon Ho; Ryu, Keon Joong

    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

  9. 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.

  10. 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.

  11. A combinatorial chemistry approach to the investigation of cerium oxide and plutonium oxide reactions with small molecules

    Science.gov (United States)

    Brady, John T.; Warner, Benjamin P.; Bridgewater, Jon S.; Havrilla, George J.; Morris, David E.; Buscher, C. Thomas

    2000-07-01

    We are currently investigating the potential chemistry of the 3013 Standard waste storage containers. These containers are filled with waste that is a mixture of inorganic salts and plutonium oxide that has been calcined to remove water and other volatiles. There has been concern about possible pressure buildup due to the formation of hydrogen or other gases. We are utilizing a combinatorial chemistry approach to investigate a range of possible reactions that may occur in the containers with various concentrations of metal oxides and inorganic salts.

  12. 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.

  13. Determination of plutonium in nitric acid solutions - Method by oxidation by cerium(IV), reduction by iron(II) ammonium sulfate and amperometric back-titration with potassium dichromate

    International Nuclear Information System (INIS)

    1987-01-01

    This International Standard specifies a precise and accurate analytical method for determining plutonium in nitric acid solutions. Plutonium is oxidized to plutonium(VI) in a 1 mol/l nitric acid solution with cerium(IV). Addition of sulfamic acid prevents nitrite-induced side reactions. The excess of cerium(IV) is reduced by adding a sodium arsenite solution, catalysed by osmium tetroxide. A slight excess of arsenite is oxidized by adding a 0.2 mol/l potassium permanganate solution. The excess of permanganate is reduced by adding a 0.1 mol/l oxalic acid solution. Iron(III) is used to catalyse the reduction. A small excess of oxalic acid does not interfere in the subsequent plutonium determination. These reduction and oxidation stages can be followed amperometrically and the plutonium is left in the hexavalent state. The sulfuric acid followed by a measured amount of standardized iron(II) ammonium sulfate solution in excess of that required to reduce the plutonium(VI) to plutonium(IV) is added. The excess iron(II) and any plutonium(III) formed to produce iron(III) and plutonium(IV) is amperometrically back-titrated using a standard potassium dichromate solution. The method is almost specifically for plutonium. It is suitable for the direct determination of plutonium in materials ranging from pure product solutions, to fast reactor fuel solutions with a uranium/plutonium ratio of up to 10:1, either before or after irradiation

  14. Grain growth kinetics in uranium-plutonium mixed oxides

    International Nuclear Information System (INIS)

    Sari, C.

    1986-01-01

    Grain growth rates were investigated in uranium-plutonium mixed oxide specimens with oxygen-to-metal ratios 1.97 and 2.0. The specimens in the form of cylindrical pellets were heated in a temperature gradient similar to that existing in a fast reactor. The results are in agreement with the cubic rate law. The mean grain size D(μm) after annealing for time t (min) is represented by D 3 -D 0 3 =1.11x10 12 . exp(-445870/RT).t and D 3 -D 0 3 =2.55x10 9 .exp(-319240/RT).t for specimens with overall oxygen-to-metal ratios 1.97 and 2.0, respectively (activation energies expressed in J/mol). An example for the influence of the oxygen-to-metal ratio on the grain growth in mixed oxide fuel during operation in a fast reactor is also given. (orig.)

  15. An assessment of the validity of cerium oxide as a surrogate for plutonium oxide gallium removal studies

    International Nuclear Information System (INIS)

    Kolman, D.G.; Park, Y.; Stan, M.; Hanrahan, R.J. Jr.; Butt, D.P.

    1999-01-01

    Methods for purifying plutonium metal have long been established. These methods use acid solutions to dissolve and concentrate the metal. However, these methods can produce significant mixed waste, that is, waste containing both radioactive and chemical hazards. The volume of waste produced from the aqueous purification of thousands of weapons would be expensive to treat and dispose. Therefore, a dry method of purification is highly desirable. Recently, a dry gallium removal research program commenced. Based on initial calculations, it appeared that a particular form of gallium (gallium suboxide, Ga 2 O) could be evaporated from plutonium oxide in the presence of a reducing agent, such as small amounts of hydrogen dry gas within an inert environment. Initial tests using ceria-based material (as a surrogate for PuO 2 ) showed that thermally-induced gallium removal (TIGR) from small samples (on the order of one gram) was indeed viable. Because of the expense and difficulty of optimizing TIGR from plutonium dioxide, TIGR optimization tests using ceria have continued. This document details the relationship between the ceria surrogate tests and those conducted using plutonia

  16. Plutonium speciation affected by environmental bacteria

    International Nuclear Information System (INIS)

    Neu, M.P.; Icopini, G.A.; Boukhalfa, H.

    2005-01-01

    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.)

  17. Redox Bias in Loss on Ignition Moisture Measurement for Relatively Pure Plutonium-Bearing Oxide Materials

    International Nuclear Information System (INIS)

    Eller, P. G.; Stakebake, J. L.; Cooper, T. D.

    2002-01-01

    This paper evaluates potential analytical bias in application of the Loss on Ignition (LOI) technique for moisture measurement to relatively pure (plutonium assay of 80 wt.% or higher) oxides containing uranium that have been stabilized according to stabilization and storage standard DOE-STD-3013-2000 (STD- 3013). An immediate application is to Rocky Flats (RF) materials derived from high-grade metal hydriding separations subsequently treated by multiple calcination cycles. Specifically evaluated are weight changes due to oxidation/reduction of multivalent impurity oxides that could mask true moisture equivalent content measurement. Process knowledge and characterization of materials representing complex-wide materials to be stabilized and packaged according to STD-3013, and particularly for the immediate RF target stream, indicate that oxides of uranium, iron and gallium are the only potential multivalent constituents expected to be present above 0.5 wt.%. The evaluation show s that of these constituents, with few exceptions, only uranium oxides can be present at a sufficient level to produce weight gain biases significant with respect to the LOI stability test. In general, these formerly high-value, high-actinide content materials are reliably identifiable by process knowledge and measurement. Significant bias also requires that UO2 components remain largely unoxidized after calcination and are largely converted to U3O8 during LOI testing at only slightly higher temperatures. Based on well-established literature, it is judged unlikely that this set of conditions will be realized in practice. We conclude that it is very likely that LOI weight gain bias will be small for the immediate target RF oxide materials containing greater than 80 wt.% plutonium plus a much smaller uranium content. Recommended tests are in progress to confirm these expectations and to provide a more authoritative basis for bounding LOI oxidation/reduction biases. LOI bias evaluation is more

  18. Ultra trace plutonium isotopic analyse by thermo ionization mass spectrometry

    International Nuclear Information System (INIS)

    Liu Xuemei; Long Kaiming; Yang Tianli

    2008-06-01

    An anion exchange combined with TTA extraction decontamination method was established. The nanogram quantity Plutonium was separated from milligram uranium and the decontamination coefficient achieved 10 7 , the recovery coefficient >95%. The active carbon powder was chosen as the ionization intensifier to increase the plutonium ion flow. Used 6.7 pg Plutonium standard as sample to measured by TIMS and the RSD of the 240 Pu/ 239 Pu ≤3.6%. (authors)

  19. Pulmonary carcinogenesis from plutonium-containing particles

    International Nuclear Information System (INIS)

    Thomas, R.G.; Smith, D.M.; Anderson, E.C.

    1980-01-01

    Induction of lung tumors by various types of radiation is of paramount concern to the nuclear industry. The data presented were obtained by exposing the pulmonary system of Syrian hamsters to particles of zirconium oxide containing various amounts of either plutonium-238 or -239 as the alpha radiation source. These particles were injected intravenously and lodged permanently in the capillary bed of the lung. When less than 20% of the lung tissue was irradiated, simulating the ''hot particle'' mode, tumors were not evident with lung burdens up to 500 nCi plutonium. More diffuse irradiation significantly increased the tumor incidence, with lung burdens of 50 to 150 nCi. When plutonium-laden microspheres were administered intratracheally, tumor production was considerably increased and the addition of 3 mg of iron oxide intratracheally further increased the incidence. Using the zirconium oxide matrix for the carrier of plutonium in aerosol particles produced tumor incidences of up to 50% in Syrian hamsters exposed by inhalation. Initial pulmonary (alveolar) burdens reached 100 nCi of plutonium. Similar inhalation studies using plutonium dioxide alone (no matrix) failed to produce any increase in lung tumorigenesis. The results are discussed in terms of possible mechanisms necessary for lung carcinogenesis. (H.K.)

  20. A kinetic study of plutonium dioxide dissolution in hydrochloric acid using iron (II) as an electron transfer catalyst

    International Nuclear Information System (INIS)

    Fife, K.W.

    1996-09-01

    Effective dissolution of plutonium dioxide has traditionally been accomplished by contact with strong nitric acid containing a small amount of fluoride at temperatures of ∼ 100 C. In spite of these aggressive conditions, PuO 2 dissolution is sometimes incomplete requiring additional contact with the solvent. This work focused on an alternative to conventional dissolution in nitric acid where an electron transfer catalyst, Fe(II), was used in hydrochloric acid. Cyclic voltammetry was employed as an in-situ analytical technique for monitoring the dissolution reaction rate. The plutonium oxide selected for this study was decomposed plutonium oxalate with > 95% of the material having a particle diameter (< 70 microm) as determined by a scanning laser microscopy technique. Attempts to dry sieve the oxide into narrow size fractions prior to dissolution in the HCl-Fe(II) solvent system failed, apparently due to significant interparticle attractive forces. Although sieve splits were obtained, subsequent scanning laser microscopy analysis of the sieve fractions indicated that particle segregation was not accomplished and the individual sieve fractions retained a particle size distribution very similar to the original powder assemblage. This phenomena was confirmed through subsequent dissolution experiments on the various screen fractions which illustrated no difference in kinetic behavior between the original oxide assemblage and the sieve fractions

  1. Plutonium ocean shipment safety between Europe and Japan

    International Nuclear Information System (INIS)

    Pierce, J.D.; Hohnstreiter, G.F.; McClure, J.D.; Smith, J.D.; Dukart, R.J.; Koski, J.A.; Braithwate, J.W.; Sorenson, N.R.; Yamamoto, K.; Kitamura, T.; Shibata, K.; Ouchi, Y.; Ito, T.

    2004-01-01

    Sandia National Laboratories (SNL) and Japan Nuclear Cycle Development Institute (JNC) have conducted an extensive emergency response planning study of the safety of the sea transport of plutonium for JNC. This study was conducted in response to international concerns about the safety of the marine transport of PuO2 powder that began with the sea transport of plutonium powder from France to Japan in 1992 using a purposebuilt ship. This emergency response planning study addressed four topics to better define the accident environment for long-range sea transport of nuclear materials. The first topic is a probabilistic safety analysis that evaluates the technical issues of transporting plutonium between Europe and Japan. An engine-room fire aboard a purposebuilt ship is evaluated as the second topic to determine the vulnerability and safety margin of radioactive material packaging for plutonium designed to meet International Atomic Energy Agency (IAEA) standards. The third topic is a corrosion study performed for generic plutonium packaging to estimate the time required to breach the containment boundary in the event of submersion in seawater. The final study topic is a worldwide survey of information on high-value cargo salvage capabilities from sunken ships. The primary purpose of this overall emergency response planning study is to describe and analyze the safety of radioactive material transportation operations for the international transportation of radioactive materials by maritime cargo vessels

  2. Plutonium ocean shipment safety between Europe and Japan

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, J D; Hohnstreiter, G F; McClure, J D; Smith, J D; Dukart, R J; Koski, J A; Braithwate, J W; Sorenson, N R [Sandia National Labs., Albuquerque, NM (United States); Yamamoto, K; Kitamura, T; Shibata, K; Ouchi, Y; Ito, T [Japan Nuclear Cycle Development Inst., Tokai-mura (Japan)

    2004-07-01

    Sandia National Laboratories (SNL) and Japan Nuclear Cycle Development Institute (JNC) have conducted an extensive emergency response planning study of the safety of the sea transport of plutonium for JNC. This study was conducted in response to international concerns about the safety of the marine transport of PuO2 powder that began with the sea transport of plutonium powder from France to Japan in 1992 using a purposebuilt ship. This emergency response planning study addressed four topics to better define the accident environment for long-range sea transport of nuclear materials. The first topic is a probabilistic safety analysis that evaluates the technical issues of transporting plutonium between Europe and Japan. An engine-room fire aboard a purposebuilt ship is evaluated as the second topic to determine the vulnerability and safety margin of radioactive material packaging for plutonium designed to meet International Atomic Energy Agency (IAEA) standards. The third topic is a corrosion study performed for generic plutonium packaging to estimate the time required to breach the containment boundary in the event of submersion in seawater. The final study topic is a worldwide survey of information on high-value cargo salvage capabilities from sunken ships. The primary purpose of this overall emergency response planning study is to describe and analyze the safety of radioactive material transportation operations for the international transportation of radioactive materials by maritime cargo vessels.

  3. Plutonium uptake by Scenedesmus obliquus as a function of isotope and oxidation state

    International Nuclear Information System (INIS)

    Tkacik, M.F.; Giesy, J.P.; Wilhite, E.L.; Corey, J.C.

    1979-01-01

    Uptake of 238 Pu 4+ , 238 Pu 6+ , 239 Pu 4+ and 239 Pu 6+ by the green alga Scenedesmus obliquus (Turp) Kutz was studied to determine whether isotope or oxidation state differences affect Pu uptake from aqueous medium by algal cells. At equivalent 238 Pu and 239 PU concentrations, even when initial oxidation states differed, accumulations of these isotopes by S.obliquus were not significantly (p>0.05) different. Plutonium accumulation by S.obliquus was log-linear. (author)

  4. Certification of a plutonium dioxide reference material for elemental analyses (EC-NRM 210)

    International Nuclear Information System (INIS)

    Le Duigou, Y.

    1990-01-01

    A new EC plutonium reference material is made available in the form of 5g samples of plutonium dioxide powder. Before weighing the material must be calcined at 1 250 0 C for two hours. The plutonium content (880.26 ± 0.44) g.kg -1 has been derived from plutonium measurements performed by three different laboratories each applying a different oxydo-reductive method. The results of the plutonium measurement, the statistical evaluation of the uncertainty of the plutonium content together with information on the impurities present in the material are given in the report

  5. Development of techniques for measuring plutonium contents in TRU wastes by NDA methods

    International Nuclear Information System (INIS)

    Matsubayashi, Toshiyuki; Kuwana, Katsumi; Morita, Tomio; Izuhara, Shigeomi; Suzuki, Masahiro

    1983-01-01

    In order to develop a technique for measuring the amount of plutonium in plutonium-contaminated (TRU) wastes, a passive gamma method was selected from many candidate methods, and examined for the suitability by applying the method to low density wastes. A segmented gamma scanner was used for the experiment. The instrument is composed mainly of a Ge(Li) detector, multichannel analyser, data processing system, turntable and transmission radiation source of (75)Se. A sham waste was prepared by adding plutonium oxide powder as a radiation source to waste matrix in a 20-1 carton box. The sham waste was put on the turntable, and the detector was set at 50 cm distance from the center of the turntable. 414 keV gamma ray emitted from (239)Pu was utilized for the assay of plutonium in the experiment. The effects of combustible (paper) waste matrix, organic chlorinated material matrix, and the distribution of plutonium source in a box on the count rate were examined, and it was concluded that 1) about 10 mg of (239)Pu contained in both matrices should be assayed by the passive gamma method, 2) 50 mg of (239) Pu was measured at 30 % confidence level with 2000 sec measuring time, 3) the effect of distribution of plutonium in a waste was able to be reduced to a value of less than 15 % by rotating the waste on the turntable. (Yoshitake, I.)

  6. Recovery of plutonium by pyroredox processing

    International Nuclear Information System (INIS)

    McNeese, J.A.; Bowersox, D.F.; Christensen, D.C.

    1985-09-01

    Using pyrochemical oxidation and reduction, we have developed a process to recover the plutonium in impure scrap with less than 95% plutonium. This plutonium metal was further purified by pyrochemical electrorefining. During development of the procedures, depleted electrorefining anodes were processed, and over 80% of the plutonium was recovered as high-purity metal in one electrorefining cycle. Over 40 kg of plutonium has been recovered from 55 kg of impure anodes with our procedures. 6 refs., 7 figs., 4 tabs

  7. Recovery of plutonium by pyroredox processing

    International Nuclear Information System (INIS)

    McNeese, J.A.; Bowersox, D.F.; Christensen, D.C.

    1985-01-01

    Using pyrochemical oxidation and reduction, we have developed a process to recover the plutonium in impure scrap with less than 95% plutonium. This plutonium metal was further purified by pyrochemical electrorefining. During development of the procedures, depleted electrorefining anodes were processed, and over 80% of the plutonium was recovered as high-purity metal in one electrorefining cycle. Over 40 kg of plutonium has been recovered from 55 kg of impure anodes with our procedures. 6 refs., 2 figs., 5 tabs

  8. Plutonium valence state distributions

    International Nuclear Information System (INIS)

    Silver, G.L.

    1974-01-01

    A calculational method for ascertaining equilibrium valence state distributions of plutonium in acid solutions as a function of the plutonium oxidation number and the solution acidity is illustrated with an example. The method may be more practical for manual use than methods based upon polynomial equations. (T.G.)

  9. Charge distribution on plutonium-containing aerosols produced in mixed-oxide reactor fuel fabrication and the laboratory

    International Nuclear Information System (INIS)

    Yeh, H.C.; Newton, G.J.; Teague, S.V.

    1976-01-01

    The inhalation toxicity of potentially toxic aerosols may be affected by the electrostatic charge on the particles. Charge may influence the deposition site during inhalation and therefore its subsequent clearance and dose patterns. The electrostatic charge distributions on plutonium-containing aerosols were measured with a miniature, parallel plate, aerosol electrical mobility spectrometer. Two aerosols were studied: a laboratory-produced 238 PuO 2 aerosol (15.8 Ci/g) and a plutonium mixed-oxide aerosol (PU-MOX, natural UO 2 plus PuO 2 , 0.02 Ci/g) formed during industrial centerless grinding of mixed-oxide reactor fuel pellets. Plutonium-238 dioxide particles produced in the laboratory exhibited a small net positive charge within a few minutes after passing through a 85 Kr discharger due to alpha particle emission removal of valence electrons. PU-MOX aerosols produced during centerless grinding showed a charge distribution essentially in Boltzmann equilibrium. The gross alpha aerosol concentrations (960-1200 nCi/l) within the glove box were sufficient to provide high ion concentrations capable of discharging the charge induced by mechanical and/or nuclear decay processes

  10. Assessments of risk following the inhalation of plutonium oxide using observed lung clearance patterns

    International Nuclear Information System (INIS)

    Ramsden, D.

    1977-10-01

    Dose commitments and risk estimates for the inhalation of plutonium oxide are calculated using the lung clearance patterns observed at AEE Winfrith. These risks are compared with published data on risks arising from a lung clearance based on the ICRP Lung Model. (author)

  11. Melting temperature of uranium - plutonium mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Tetsuya; Hirosawa, Takashi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960`s and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960`s and that some of the 1960`s data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO{sub 2} - PuO{sub 2} - PuO{sub 1.61} ideal solution model, and then formulized. (J.P.N.)

  12. Melting temperature of uranium - plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Ishii, Tetsuya; Hirosawa, Takashi

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960's and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960's and that some of the 1960's data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO 2 - PuO 2 - PuO 1.61 ideal solution model, and then formulized. (J.P.N.)

  13. The use of plutonium in Swedish reactors

    International Nuclear Information System (INIS)

    Forsstroem, H.

    1982-09-01

    The report deals with the utilization of plutonium in Swedish nuclear power plants. The plutonium content of the mixed oxide fuel will normally be 3-7 per cent. The processing of spent nuclear fuel will produce about 6 ton plutonium. The use of mixed oxide fuel in Forsmark 3 and Oskarshamn 3 is discussed. The fuel cycle will start with the manufacturing of the fuel elements abroad and proceeds with transport and utilization, storing of spent fuel about 40 years in Sweden followed by direct disposal. The manufacture and use of mixed oxide (MOX) fuel is based on well-known techniques. Approximately 20 000 MOX fuel rods have been irradiated and the fuel is essentially equivalent to uranium oxide fuel. 30-50 per cent of the core may be composed of MOX-fuel without any effect on the operation and safety of the reactor which has been originally designed for uranium fuel. The evaluation of international fuel cycle (INFCE) states that the proliferation risks are very small. The recycling of plutonium will reduce demand for enriched uranium and the calculations show that 6.3 ton plutonium will replace the enrichment of 600 ton natural uranium. (G.B.)

  14. Oxidation-state distribution of plutonium in surface and subsurface waters at Thule, northwest Greenland

    International Nuclear Information System (INIS)

    McMahon, C.A.; Leon Vintro, L.; Mitchell, P.I.; Dahlgaard, H.

    2000-01-01

    The speciation of plutonium in Arctic waters sampled on the northwest Greenland shelf in August 1997 is discussed in this paper. Specifically, we report the results of analyses carried out on seawater sampled (a) close to the Thule air base where, in 1968, a US military aircraft carrying four nuclear weapons crashed on sea ice, releasing kilogram quantities of plutonium to the snow pack and underlying seabed sediments, and (b) at a reference station (Upernavik) located ∼400 km to the south. The data show that most of the plutonium in the dissolved phase at Thule is in the form of Pu(V, VI) (mean: 68±6%; n=6), with little if any distinction apparent between surface and bottom waters. Further, the oxidation state distribution at stations close to the accident site is similar to that measured at Upernavik, remote from this site. It is also similar to the distribution observed in shelf waters at mid-latitudes, suggesting that the underlying processes controlling plutonium speciation are insensitive to temperature over the range 0-25 deg. C. Measurements using tangential-flow ultrafiltration indicate that virtually all of the plutonium (including the fraction in a reduced chemical form) is present as fully dissolved species. Most of this plutonium would seem to be of weapons fallout origin, as the mean 238 Pu/ 239,240 Pu activity ratio in the water column (dissolved phase) at Thule (0.06±0.02; n=10) is similar to the global fallout ratio at this latitude (∼0.04). Thus, there is little evidence of weapons-grade plutonium in the water column at Thule at the present time

  15. Nondestructive assay of plutonium in empty stainless steel boxes by apparent mass method

    International Nuclear Information System (INIS)

    Agarwal, C.; Chaudhury, S.; Nathaniel, T.N.; Goswami, A.

    2012-01-01

    Apparent mass method (Venkataraman and Croft, Nucl Instrum Methods Phys Res A 505:527, 2003), initially standardized for the assay of Pu (Agarwal et al., J Nucl Mater 651:386, 2007) has been used to get Pu amount in empty stainless steel boxes generally used for storing and transferring plutonium oxide powders. The results have been compared with the neutron coincidence counting results and have been found to match well. The advantage of the method is that it can be used for any sample with nonstandard geometry and with uncertain source distribution. (author)

  16. Polarographic study of the electrochemical properties of plutonium; Etude polarographique des proprietes electrochimiques du plutonium

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-07-01

    The behaviour of the different degrees of oxidation of plutonium have been studied using a falling drop mercury electrode in a non-complexing medium: dilute perchloric acid. In this medium it is possible to carry out the polarographic determination of plutonium using the reduction step situated at - 0.54 V/ECS which corresponds to the passage from the degree of oxidation(V) to the degree(III). The modifications brought about by a complexing ion, acetate, are then observed and interpreted. The existence of two plutonium(IV) acetic complexes has been shown; one is a polymerized substance. (author) [French] Le comportement des differents degres d'oxydation du plutonium est etudie a l'electrode a gouttes de mercure dans un milieu non complexant: l'acide perchlorique dilue. Il est possible dans ce milieu d'effectuer le dosage polarographique du plutonium en utilisant la vague de reduction situee a - 0.54 V/ECS qui correspond au passage du degre d'oxydation(V) au degre(III). Les modifications apportees par un ion complexant, l'acetate, sont ensuite observees et interpretees. Deux complexes acetiques du plutonium(IV) ont ete mis en evidence, dont l'un est un compose polymerise. (auteur)

  17. Oxidation-state distribution of plutonium in surface and subsurface waters at Thule, northwest Greenland

    DEFF Research Database (Denmark)

    McMahon, C.A.; Vintró, L.L.; Mitchell, P.I.

    2000-01-01

    (V, VI) (mean, 68 +/- 6%; n = 6), with little if any distinction apparent between surface and bottom waters. Further, the oxidation state distribution at stations close to the accident site is similar to that measured at Upernavik, remote from this site. It is also similar to the distribution observed...... in shelf waters at midlatitudes, suggesting that the underlying processes controlling plutonium speciation are insensitive to temperature over the range 0-25 degrees C. Measurements using tangential-flow ultrafiltration indicate that virtually all of the plutonium (including the fraction in a reduced...... chemical form) is present as fully dissolved species. Most of this plutonium would seem to be of weapons fallout origin, as the mean Pu-238/Pu-239,Pu-240 activity ratio in the water column (dissolved phase) at Thule (0.06 +/- 0.02; n = 10) is similar to the global fallout ratio at this latitude...

  18. Effect of mixing state on criticality safety evaluation in MOX powder and additive

    International Nuclear Information System (INIS)

    Yamamoto, Toshihiro; Miyoshi, Yoshinori

    2005-01-01

    Criticality safety analyses are discussed in which MOX powder and additive (e.g. zinc-stearate) are mixed in a powder treatment process of MOX fuel fabrication. The multiplication factor k eff is largely affected by how they are mixed, i.e., how the density and volume change with the mixing. In general, k eff increases when MOX powder is mixed with zinc-stearate. However, plutonium content and density of MOX powder make a difference in the k eff 's changes. Especially, MOX powder with a higher plutonium content and a higher density is not always unsafe in terms of criticality if it is mixed with zinc-stearate. (author)

  19. Studies involving direct heating of uranium and plutonium oxides by microwaves

    Energy Technology Data Exchange (ETDEWEB)

    Mallik, G K; Malav, R K; Karande, A P; Bhargava, V K; Kamath, H S [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility

    1997-08-01

    Nuclear fuel fabrication and recovery of nuclear materials from scraps generated during fabrication involve heating steps like dewaxing, sintering, roasting of scraps, calcination, etc. The dielectric properties of uranium and plutonium oxides place them in the category of materials which are susceptible to absorption of microwaves. The studies were carried out to explore the microwave heating technique for these steps required in nuclear fuel fabrication and scrap recovery laboratories. (author). 1 ref.

  20. Status of plutonium recycle from mixed oxide fuel fabrication wastes (U,Pu)O2 facility activities

    International Nuclear Information System (INIS)

    Quesada, Calixto A.; Adelfang, Pablo; Greiner, G.; Orlando, Oscar S.; Mathot, Sergio R.

    1999-01-01

    Within the specific subject of mixed oxides corresponding to the Fuel Cycle activities performed at CNEA, the recovery of plutonium from wastes originated during tests and pre-fabrication stages is performed. (author)

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

    International Nuclear Information System (INIS)

    Menghini, Jorge E.; Marchi, Daniel E.; Orosco, Edgardo H.; Greco, Luis

    2000-01-01

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

  2. Test of a sample container for shipment of small size plutonium samples with PAT-2

    International Nuclear Information System (INIS)

    Kuhn, E.; Aigner, H.; Deron, S.

    1981-11-01

    A light-weight container for the air transport of plutonium, to be designated PAT-2, has been developed in the USA and is presently undergoing licensing. The very limited effective space for bearing plutonium required the design of small size sample canisters to meet the needs of international safeguards for the shipment of plutonium samples. The applicability of a small canister for the sampling of small size powder and solution samples has been tested in an intralaboratory experiment. The results of the experiment, based on the concept of pre-weighed samples, show that the tested canister can successfully be used for the sampling of small size PuO 2 -powder samples of homogeneous source material, as well as for dried aliquands of plutonium nitrate solutions. (author)

  3. Storage stability of cauliflower soup powder: The effect of lipid oxidation and protein degradation reactions.

    Science.gov (United States)

    Raitio, Riikka; Orlien, Vibeke; Skibsted, Leif H

    2011-09-15

    Soups based on cauliflower soup powders, prepared by dry mixing of ingredients and rapeseed oil, showed a decrease in quality, as evaluated by a sensory panel, during the storage of the soup powder in the dark for up to 12weeks under mildly accelerated conditions of 40°C and 75% relative humidity. Antioxidant, shown to be effective in protecting the rapeseed bulk oil, used for the powder preparation, had no effect on storage stability of the soup powder. The freshly prepared soup powder had a relatively high concentration of free radicals, as measured by electron spin resonance spectroscopy, which decreased during storage, and most remarkably during the first two weeks of storage, with only marginal increase in lipid hydroperoxides as primary lipid oxidation products, and without any increase in secondary lipid oxidation products. Analyses of volatiles by SPME-GC-MS revealed a significant increase in concentrations of 2-methyl- and 3-methyl butanals, related to Maillard reactions, together with an increase in 2-acetylpyrrole concentration. The soup powders became more brown during storage, as indicated by a decreasing Hunter L-value, in accord with non-enzymatic browning reactions. A significant increase in the concentrations of dimethyl disulfide in soup powder headspace indicated free radical-initiated protein oxidation. Protein degradation, including Maillard reactions and protein oxidation, is concluded to be more important than lipid oxidation in determining the shelf-life of dry cauliflower soup powder. Copyright © 2011 Elsevier Ltd. All rights reserved.

  4. NMR study of nanophase Al/Al-oxide powder and consolidated composites

    International Nuclear Information System (INIS)

    Suits, B.H.; Apte, P.; Wilken, D.E.; Siegel, R.W.

    1994-10-01

    27 Al Nuclear Magnetic Resonance (NMR) measurements from aluminum powders and consolidated nanophase aluminum made from those powders are presented. The signals from the metal and surface oxidation are easily separated and are compared before and after consolidation. The results presented indicate that the oxide coating becomes the interface region within the nanophase composite material and that during consolidation the metal has undergone a deformation equivalent to that seen for bulk material under a compressive strain of between 4% and 8%

  5. Microstructural evolution of ferritic steel powder during mechanical alloying with iron oxide

    Energy Technology Data Exchange (ETDEWEB)

    Wen, Yuren; Liu, Yong; Liu, Donghua; Tang, Bei [Central South Univ., State Key Lab. of Powder Metallurgy, Changsha (China); Liu, C.T. [The Hong Kong Polytechnic Univ., Dept. of Mechanical Engineering, Hong Kong (China)

    2011-02-15

    Mechanical alloying of mixed powders is of great importance for preparing oxide dispersion strengthened ferritic steels. In this study, the microstructural evolution of ferritic steel powder mixed with TiH{sub x}, YH{sub 2} and Fe{sub 2}O{sub 3} in the process of mechanical alloying is systematically investigated by using X-ray diffraction analysis, scanning electron microscopy, transmission electron microscopy and microhardness tests. It is found that titanium, yttrium hydrides and iron oxide are completely dissolved during milling, and homogeneous element distribution can be achieved after milling for 12 h. The disintegration of the composite powder particles occurs at 24 h and reaches the balance of welding and fracturing after 36 h. The oxygen content increases sharply with the disintegration of powder particles due to the absorption of oxygen at the solid/gas interface from the milling atmosphere, which is the main source of extra oxygen in the milled powder. Grain refinement down to nanometer level occurs due to the severe plastic deformation of particles; however, the grain size does not change much with further disintegration of particles. The hardness increases with milling time and then becomes stable during further milling. The study indicates that the addition of iron oxide and hydrides may be more beneficial for the dispersion and homogenization of chemical compositions in the powder mixture, thus shortening the mechanical alloying process. (orig.)

  6. The economics of plutonium recycle

    International Nuclear Information System (INIS)

    James, R.A.

    1977-11-01

    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)

  7. Safe disposal of surplus plutonium

    Science.gov (United States)

    Gong, W. L.; Naz, S.; Lutze, W.; Busch, R.; Prinja, A.; Stoll, W.

    2001-06-01

    About 150 tons of weapons grade and weapons usable plutonium (metal, oxide, and in residues) have been declared surplus in the USA and Russia. Both countries plan to convert the metal and oxide into mixed oxide fuel for nuclear power reactors. Russia has not yet decided what to do with the residues. The US will convert residues into a ceramic, which will then be over-poured with highly radioactive borosilicate glass. The radioactive glass is meant to provide a deterrent to recovery of plutonium, as required by a US standard. Here we show a waste form for plutonium residues, zirconia/boron carbide (ZrO 2/B 4C), with an unprecedented combination of properties: a single, radiation-resistant, and chemically durable phase contains the residues; billion-year-old natural analogs are available; and criticality safety is given under all conceivable disposal conditions. ZrO 2/B 4C can be disposed of directly, without further processing, making it attractive to all countries facing the task of plutonium disposal. The US standard for protection against recovery can be met by disposal of the waste form together with used reactor fuel.

  8. Calorimetric systems designed for in-field nondestructive assay of plutonium-bearing materials

    International Nuclear Information System (INIS)

    Roche, C.T.; Perry, R.B.; Lewis, R.N.; Jung, E.A.; Haumann, J.R.

    1978-01-01

    Conventional calorimetric design measures the temperature rise of a plutonium-containing sample chamber in contact with a large water-bath heat sink. This design lacks the mobility needed by inspection personnel. The Argonne National Laboratory air-chamber isothermal calorimeters are low-thermal capacitance devices which eliminate the need for large, constant-temperature heat sinks. A bulk calorimeter designed to measure sealed containers holding up to 3 kg Pu, and a small-sample calorimeter designed to measure mixed-oxide fuel pellets and powders are discussed. The operational characteristics of these instruments are described, and the results of sample assays are presented

  9. SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

    Science.gov (United States)

    Spence, R.; Lister, M.W.

    1958-12-16

    Uranium and plutonium can be separated from neutron-lrradiated uranium by a process consisting of dissolvlng the lrradiated material in nitric acid, saturating the solution with a nitrate salt such as ammonium nitrate, rendering the solution substantially neutral with a base such as ammonia, adding a reducing agent such as hydroxylamine to change plutonium to the trivalent state, treating the solution with a substantially water immiscible organic solvent such as dibutoxy diethylether to selectively extract the uranium, maklng the residual aqueous solutlon acid with nitric acid, adding an oxidizing agent such as ammonlum bromate to oxidize the plutonium to the hexavalent state, and selectlvely extracting the plutonium by means of an immlscible solvent, such as dibutoxy dlethyletber.

  10. Plutonium roundtable discussion

    International Nuclear Information System (INIS)

    Penneman, R.A.

    1982-01-01

    The roundtable discussion began with remarks by the chairman who pointed out the complicated nature of plutonium chemistry. Judging from the papers presented at this symposium, he noticed a pattern which indicated to him the result of diminished funding for investigation of basic plutonium chemistry and funding focused on certain problem areas. Dr. G.L. silver pointed to plutonium chemists' erroneous use of a simplified summary equation involving the disproportionation of Pu(EV) and their each of appreciation of alpha coefficients. To his appreciation of alpha coefficients. To his charges, Dr. J.T. Bell spoke in defense of the chemists. This discussion was followed by W.W. Schulz's comments on the need for experimental work to determine solubility data for plutonium in its various oxidation states under geologic repository conditions. Discussion then turned to plutonium pyrachemical process with Dana C. Christensen as the main speaker. This paper presents edited versions of participants' written version

  11. Plutonium-induced lymphadenitis in beagles

    International Nuclear Information System (INIS)

    Dagle, G.E.; Park, J.F.

    1976-01-01

    Plutonium oxide particles accumulate in the tracheobronchial lymph nodes of beagles exposed by inhalation and in the popliteal lymph nodes after subcutaneous injection in the hind paws. The sequence of histopathologic changes after phagocytosis of particles included necrosis, increased numbers of macrophages, and fibroplasia. Scar tissue eventually replaced the normal architecture of the lymph nodes and sequestered the plutonium particles from surrounding parenchyma. Electron microscopy showed that plutonium particles were aggregated in phagolysosomes of macrophages

  12. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    International Nuclear Information System (INIS)

    Eccleston, G.W.; Menlove, H.O.; Abhold, M.; Baker, M.; Pecos, J.

    1999-01-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency 3 He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the 240 Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual

  13. Comparison of predicted with observed biokinetics of inhaled plutonium nitrate and gadolinium oxide in humans

    International Nuclear Information System (INIS)

    Hodgson, A.; Shutt, A.L.; Etherington, G.; Hodgson, S.A.; Rance, E.; Stradling, G.N.; Youngman, M.J.; Ziesenis, A.; Kreyling, W.G.

    2003-01-01

    The absorption kinetics to blood of plutonium and gadolinium after inhalation as nitrate and oxide in humans and animals has been studied. For each material, values describing the time dependence of absorption were derived from the studies in animals and used with the ICRP human respiratory tract model to predict lung retention and cumulative amounts to blood for the volunteers inhaling the same materials. Comparison with the observed behaviour in the volunteers suggests that absorption of plutonium and gadolinium is reasonably species independent, and that data obtained from animal studies can be used to assess their biokinetic behaviour in humans. (author)

  14. Fabrication of Cerium Oxide and Uranium Oxide Microspheres for Space Nuclear Power Applications

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey A. Katalenich; Michael R. Hartman; Robert C. O' Brien

    2013-02-01

    Cerium oxide and uranium oxide microspheres are being produced via an internal gelation sol-gel method to investigate alternative fabrication routes for space nuclear fuels. Depleted uranium and non-radioactive cerium are being utilized as surrogates for plutonium-238 (Pu-238) used in radioisotope thermoelectric generators and for enriched uranium required by nuclear thermal rockets. While current methods used to produce Pu-238 fuels at Los Alamos National Laboratory (LANL) involve the generation of fine powders that pose a respiratory hazard and have a propensity to contaminate glove boxes, the sol-gel route allows for the generation of oxide microsphere fuels through an aqueous route. The sol-gel method does not generate fine powders and may require fewer processing steps than the LANL method with less operator handling. High-quality cerium dioxide microspheres have been fabricated in the desired size range and equipment is being prepared to establish a uranium dioxide microsphere production capability.

  15. Skeletal lesions from inhaled plutonium in beagles

    International Nuclear Information System (INIS)

    Dagle, G.E.; Park, J.F.; Weller, R.E.; Ragan, H.A.; McClanahan, B.J.; Fisher, D.R.

    1984-10-01

    The report briefly reviews the skeletal effects observed in ongoing lifespan studies in beagle dogs at 13, 10, and 7 years, respectively, after inhalation exposure to 239 Pu oxide and nitrate or 238 Pu oxide. Plutonium nitrate was chosen to represent soluble material more readily translocated to bone and other tissues than the oxide. Bone lesions related to plutonium exposure were observed only in dogs exposed to 238 Pu oxide and 239 Pu nitrate. The skeleton accumulated approximately 2% ( 239 Pu oxide), 45% ( 238 Pu oxide) or 50% ( 239 Pu nitrate) of the final body burdens at 13, 10, and 7 years, respectively, after exposure. 11 references, 2 figures

  16. An isotopic analysis system for plutonium samples enriched in 238Pu

    International Nuclear Information System (INIS)

    Ruhter, W.D.; Camp, D.C.

    1991-08-01

    We have designed and built a gamma-ray spectrometer system that measures the relative plutonium isotopic abundances of plutonium oxide enriched in 238 Pu. The first system installed at Westinghouse Savannah River Company was tested and evaluated on plutonium oxide in stainless steel EP60/61 containers. 238 Pu enrichments ranged from 20% to 85%. Results show that 200 grams of plutonium oxide in an EP60.61 container can be measured with ±0.3% precision and better than ±1.0% accuracy in the specific power using a counting time of 50 minutes. 3 refs., 2 figs

  17. Cleaning up the Legacy of the Cold War: Plutonium Oxides and the Role of Synchrotron Radiation Research

    Energy Technology Data Exchange (ETDEWEB)

    Clark, David Lewis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-21

    The deceptively simple binary formula of AnO2 belies an incredibly complex structural nature, and propensity to form mixed-valent, nonstoichiometric phases of composition AnO2±x. For plutonium, the very formation of PuO2+x has challenged a long-established dogma, and raised fundamental questions for long-term storage and environmental migration. This presentation covers two aspects of Los Alamos synchrotron radiation studies of plutonium oxides: (1) the structural chemistry of laboratory-prepared AnO2+x systems (An = U, Pu; 0 ≤ x ≤ 0.25) determined through a combination of x-ray absorption fine structure spectroscopy (XAFS) and x-ray scattering of laboratory prepared samples; and (2) the application of synchrotron radiation towards the decontamination and decommissioning of the Rocky Flats Environmental Technology Site. Making the case for particle transport mechanisms as the basis of plutonium and americium mobility, rather than aqueous sorption-desorption processes, established a successful scientific basis for the dominance of physical transport processes by wind and water. The scientific basis was successful because it was in agreement with general theory on insolubility of PuO2 in oxidation state IV, results of ultrafiltration analyses of field water/sediment samples, XAFS analyses of soil, sediment, and concrete samples, and was also in general agreement with on-site monitoring data. This understanding allowed Site contractors to rapidly move to application of soil erosion and sediment transport models as the means of predicting plutonium and americium transport, which led to design and application of site-wide soil erosion control technology to help control downstream concentrations of plutonium and americium in streamflow.

  18. Cleaning up the Legacy of the Cold War: Plutonium Oxides and the Role of Synchrotron Radiation Research

    International Nuclear Information System (INIS)

    Clark, David Lewis

    2015-01-01

    The deceptively simple binary formula of AnO 2 belies an incredibly complex structural nature, and propensity to form mixed-valent, nonstoichiometric phases of composition AnO 2±x . For plutonium, the very formation of PuO 2+x has challenged a long-established dogma, and raised fundamental questions for long-term storage and environmental migration. This presentation covers two aspects of Los Alamos synchrotron radiation studies of plutonium oxides: (1) the structural chemistry of laboratory-prepared AnO 2+x systems (An = U, Pu; 0 ≤ x ≤ 0.25) determined through a combination of x-ray absorption fine structure spectroscopy (XAFS) and x-ray scattering of laboratory prepared samples; and (2) the application of synchrotron radiation towards the decontamination and decommissioning of the Rocky Flats Environmental Technology Site. Making the case for particle transport mechanisms as the basis of plutonium and americium mobility, rather than aqueous sorption-desorption processes, established a successful scientific basis for the dominance of physical transport processes by wind and water. The scientific basis was successful because it was in agreement with general theory on insolubility of PuO 2 in oxidation state IV, results of ultrafiltration analyses of field water/sediment samples, XAFS analyses of soil, sediment, and concrete samples, and was also in general agreement with on-site monitoring data. This understanding allowed Site contractors to rapidly move to application of soil erosion and sediment transport models as the means of predicting plutonium and americium transport, which led to design and application of site-wide soil erosion control technology to help control downstream concentrations of plutonium and americium in streamflow.

  19. Electrochemistry of plutonium in molten halides

    International Nuclear Information System (INIS)

    McCurry, L.E.; Moy, G.M.M.; Bowersox, D.F.

    1987-01-01

    The electrochemistry of plutonium in molten halides is of technological importance as a method of purification of plutonium. Previous authors have reported that plutonium can be purified by electrorefining impure plutonium in various molten haldies. Work to eluciate the mechanism of the plutonium reduction in molten halides has been limited to a chronopotentiometric study in LiCl-KCl. Potentiometric studies have been carried out to determine the standard reduction potential for the plutonium (III) couple in various molten alkali metal halides. Initial cyclic voltammetric experiments were performed in molten KCL at 1100 K. A silver/silver chloride (10 mole %) in equimolar NaCl-KCl was used as a reference electrode. Working and counter electrodes were tungsten. The cell components and melt were contained in a quartz crucible. Background cyclic voltammograms of the KCl melt at the tungsten electrode showed no evidence of electroactive impurities in the melt. Plutonium was added to the melt as PuCl/sub 3/, which was prepared by chlorination of the oxide. At low concentrations of PuCl/sub 3/ in the melt (0.01-0.03 molar), no reduction wave due to the reduction of Pu(III) was observed in the voltammograms up to the potassium reduction limit of the melt. However on scan reversal after scanning into the potassium reduction limit a new oxidation wave was observed

  20. Speciation of the oxidation states of plutonium in aqueous solutions by UV/Vis spectroscopy, CE-ICP-MS and CE-RIMS

    International Nuclear Information System (INIS)

    Buerger, S.; Banik, N.L.; Buda, R.A.; Kratz, J.V.; Kuczewski, B.; Trautmann, N.

    2007-01-01

    For the speciation of the plutonium oxidation states in aqueous solutions, the online coupling of capillary electrophoresis (CE) with inductively coupled plasma mass spectrometry (ICP-MS) has been developed. Depending on the radius/electrical charge ratio, the oxidation states III, IV, V, and VI of plutonium are separated by CE, based on the different migration times through the capillary and are detected by ICP-MS. The detection limit is 20 ppb, i.e. 10 9 -10 10 atoms (10 -12 -10 -13 g) for one oxidation state with an uncertainty of the reproducibility of the migration times of ≤ 1% and ≤ 5% for the peak area. The redox kinetics of the different plutonium oxidation states in the presence of humic substances (humic and fulvic acid) have been studied. A relatively rapid reduction of Pu(VI) (10 to 1000 h) in contact with Gorleben fulvic or Aldrich humic acid could be observed, depending on the pH of the solution. Furthermore, at pH=1, a reduction to Pu(III) and Pu(IV) in a mixture of all four oxidation states in contact with Gorleben fulvic acid after one month has been observed. In order to improve the sensitivity of the CE method, the offline coupling of CE to resonance ionization mass spectrometry (RIMS) has been explored. First applications of this new speciation method are presented. (orig.)

  1. Flexible process options for the immobilisation of residues and wastes containing plutonium

    International Nuclear Information System (INIS)

    Stewart, M.W.A.; Moricca, S.A.; Day, R. A.; Begg, B. D.; Scales, C. R.; Maddrell, E. R.; Eilbeck, A. B.

    2007-01-01

    Residues and waste streams containing plutonium present unique technical, safety, regulatory, security, and socio-political challenges. In the UK these streams range from lightly plutonium contaminated materials (PCM) through to residue s resulting directly from Pu processing operations. In addition there are potentially stocks of Pu oxide powders whose future designation may be either a waste or an asset, due to their levels of contamination making their reuse uneconomic, or to changes in nuclear policy. While waste management routes exist for PCM, an immobilisation process is required for streams containing higher levels of Pu. Such a process is being developed by Nexia Solutions and ANSTO to treat and immobilise Pu waste and residues currently stored on the Sellafield site. The characteristics of these Pu waste streams are highly variable. The physical form of the Pu waste ranges from liquids, sludges, powders/granules, to solid components (e.g., test fuels), with the Pu present as an ion in solution, as a salt, metal, oxide or other compound. The chemistry of the Pu waste streams also varies considerably with a variety of impurities present in many waste streams. Furthermore, with fissile isotopes present, criticality is an issue during operations and in the store or repository. Safeguards and security concerns must be assessed and controlled. The process under development, by using a combination of tailored waste form chemistry combined with flexible process technology aims to develop a process line to handle a broad range of Pu waste streams. It aims to be capable of dealing with not only current arisings but those anticipated to arise as a result of future operations or policy changes. (authors)

  2. Comments on the note by L.S. Nelson: ''unusually high (oxidizer/Pu) ratios in the macro-residues from plutonium droplet combustion''

    International Nuclear Information System (INIS)

    Allen, M.D.; Morgan, L.G.

    1982-01-01

    An alternative explanation for unusually high ratios of oxidizers/plutonium in the residues from plutonium droplet combustion is offered in response to a note by L.S. Nelson. Two additional experiments are suggested that would be helpful in clarifying the high oxidizer/Pu ratios. The first would be to measure the true surface area for a spherule after its explosion. The second is an experiment analogous to a gas analysis experiment. The conclusion is put forth that adsorbed gases are the most likely cause of the high O/M ratios described by Nelson

  3. Plutonium spectrophotometric analysis

    International Nuclear Information System (INIS)

    Esteban, A.; Cassaniti, P.; Orosco, E.H.

    1990-01-01

    Plutonium ions in solution have absorption spectra so different that it is possible to use them for analytical purposes. Detailed studies have been performed in nitric solutions. Some very convenient methods for the determination of plutonium and its oxidation states, especially the ratios Pu(III):Pu(IV) and Pu(IV):Pu(VI) in a mixture of both, have been developed. These methods are described in this paper, including: a) Absorption spectra for plutonium (III), (IV), (VI) and mixtures. b) Relative extinction coefficients for the above mentioned species. c) Dependences of the relative extinction coefficients on the nitric acid concentration and the plutonium VI deviation from the Beer-Lambert law. The developed methods are simple and rapid and then, suitable in process control. Accuracy is improved when relative absorbance measurements are performed or controlled the variables which have effect on the spectra and extinction coefficients. (Author) [es

  4. Homogeneity of blended nuclear fuel powders after pneumatic transport

    International Nuclear Information System (INIS)

    Smeltzer, E.E.; Skriba, M.C.; Lyon, W.L.

    1982-01-01

    A study of the pneumatic transport of fine (approx. 1μm) cohesive nuclear fuel powders was conducted for the U.S. Department of Energy to demonstrate the feasibility of this method of transport and to develop a design data base for use in a large scale nuclear fuel production facility. As part of this program, a considerable effort was directed at following the homogeneity of blended powders. Since different reactors require different enrichments, blending and subsequent transport are critical parts of the fabrication sequence. The various materials used represented analogs of a wide range of powders and blends that could be expected in a commercial mixed oxide fabrication facility. All UO 2 powders used were depleted and a co-precipitated master mix of (U, Th)O 2 was made specifically for this program, using thorium as an analog for plutonium. In order to determine the effect of pneumatic transport on a blended powder, samples were taken from a feeder vessel before each test, and from a receiver vessel and a few line sections after each transfer test. The average difference between the before and after degree of non-homogeneity was < 1%, for the 21 tests considered. This shows that overall, the pneumatic transport of blended, fine nuclear fuel powders is possible, with only minor unblending occurring

  5. Study on Exothermic Oxidation of Acrylonitrile-butadiene-styrene (ABS Resin Powder with Application to ABS Processing Safety

    Directory of Open Access Journals (Sweden)

    Jenq-Renn Chen

    2010-08-01

    Full Text Available Oxidative degradation of commercial grade ABS (Acrylonitrile-butadiene-styrene resin powders was studied by thermal analysis. The instabilities of ABS containing different polybutadiene (PB contents with respect to temperature were studied by Differential Scanning Calorimeter (DSC. Thermograms of isothermal test and dynamic scanning were performed. Three exothermic peaks were observed and related to auto-oxidation, degradation and oxidative decomposition, respectively. Onset temperature of the auto-oxidation was determined to be around 193 °C. However, threshold temperature of oxidation was found to be as low as 140 °C by DSC isothermal testing. Another scan of the powder after degeneration in air showed an onset temperature of 127 °C. Reactive hazards of ABS powders were verified to be the exothermic oxidation of unsaturated PB domains, not the SAN (poly(styrene-acrylonitrile matrix. Heat of oxidation was first determined to be 2,800 ± 40 J per gram of ABS or 4,720 ± 20 J per gram of PB. Thermal hazards of processing ABS powder are assessed by adiabatic temperature rise at process conditions. IR spectroscopy associated with heat of oxidation verified the oxidative mechanism, and these evidences excluded the heat source from the degradation of SAN. A specially prepared powder of ABS without adding anti-oxidant was analyzed by DSC for comparing the exothermic behaviors. Exothermic onset temperatures were determined to be 120 °C and 80 °C by dynamic scanning and isothermal test, respectively. The assessment successfully explained fires and explosions in an ABS powder dryer and an ABS extruder.

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

    International Nuclear Information System (INIS)

    Lagrave, H.; Beretti, C.; Bros, P.

    2008-01-01

    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)

  7. 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)

  8. Design-Only Conceptual Design Report: Plutonium Immobilization Plant

    International Nuclear Information System (INIS)

    DiSabatino, A.; Loftus, D.

    1999-01-01

    This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant

  9. Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 3

    International Nuclear Information System (INIS)

    1976-08-01

    An assessment is presented of the health, safety and environmental effects of the entire light water reactor fuel cycle, considering the comparative effects of three major alternatives: no recycle, recycle of uranium only, and recycle of both uranium and plutonium. The assessment covers the period from 1975 through the year 2000 and includes the cumulative effects for the entire period as well as projections for specific years. Topics discussed include: the light water reactor with plutonium recycle; mixed oxide fuel fabrication; reprocessing plant operations; supporting uranium fuel cycle; transportation of radioactive materials; radioactive waste management; storage of plutonium; radiological health assessment; extended spent fuel storage; and blending of plutonium and uranium at reprocessing plants

  10. Determination of uranium and plutonium by sequential potentiometric titration

    International Nuclear Information System (INIS)

    Kato, Yoshiharu; Takahashi, Masao

    1976-01-01

    The determination of uranium and plutonium in mixed oxide fuels has been developed by sequential potentiometric titration. A weighed sample of uranium and plutonium oxides is dissolved in a mixture of nitric and hydrofluoric acids and the solution is fumed with sulfuric acid. After the reduction of uranium and plutonium to uranium(IV) and plutonium(III) by chromium(II) sulfate, 5 ml of buffer solution (KCl-HCl, pH 1.0) is added to the solution. Then the solution is diluted to 30 ml with water and the pH of the solution is adjusted to 1.0 -- 1.5 with 1 M sodium hydroxide. The solution is stirred until the oxidation of the excess of chromium(II) by air is completed. After the removal of dissolved oxygen by bubbling nitrogen through the solution for 10 minutes, uranium (IV) is titrated with 0.1 N cerium(IV) sulfate. Then, plutonium is titrated by 0.02 N cerium(IV) sulfate. When a mixture of uranium and plutonium is titrated with 0.1 N potassium dichromate potentiometrically, the potential change at the end point of plutonium is very small and the end point of uranium is also unclear when 0.1 N potassium permanganate is used as a titrant. In the present method, nitrate, fluoride and copper(II) interfere with the determination of plutonium and uranium. Iron interferes quantitatively with the determination of plutonium but not of uranium. Results obtained in applying the proposed method to 50 mg of mixtures of plutonium and uranium ((7.5 -- 50))% Pu were accurate to within 0.15 mg of each element. (auth.)

  11. United Kingdom experience in plutonium transportation

    International Nuclear Information System (INIS)

    1978-11-01

    This paper describes the extensive experience of the UKAEA and British Nuclear Fuels Limited in transporting plutonium within the UK over the last 20 years and to destinations overseas since the early 1960s. The aspects covered include: the form of plutonium transported (nitrate, oxide, mixed oxide etc.); UK and international regulations (e.g. covering safety and safeguards matters); insurance; container design; mode of transport; physical protection; and a safety analysis. Costs are estimated to be largely independent of quantity

  12. An autoradiographical method using an imaging plate for the analyses of plutonium contamination in a plutonium handling facility

    International Nuclear Information System (INIS)

    Takasaki, Koji; Sagawa, Naoki; Kurosawa, Shigeyuki; Mizuniwa, Harumi

    2011-01-01

    An autoradiographical method using an imaging plate (IP) was developed to analyze plutonium contamination in a plutonium handling facility. The IPs were exposed to ten specimens having a single plutonium particle. Photostimulated luminescence (PSL) images of the specimens were taken using a laser scanning machine. One relatively large spot induced by α-radioactivity from plutonium was observed in each PSL image. The plutonium-induced spots were discriminated by a threshold derived from background and the size of the spot. A good relationship between the PSL intensities of the spots and α-radioactivities measured using a radiation counter was obtained by least-square fitting, taking the fading effect into consideration. This method was applied to workplace monitoring in an actual uranium-plutonium mixed oxide (MOX) fuel fabrication facility. Plutonium contaminations were analyzed in ten other specimens having more than two plutonium spots. The α-radioactivities of plutonium contamination were derived from the PSL images and their relative errors were evaluated from exposure time. (author)

  13. Technological alternatives for plutonium transport

    International Nuclear Information System (INIS)

    1978-12-01

    This paper considers alternative transport modes (air, sea, road, rail) for moving (1) plutonium from a reprocessing plant to a store or a fuel fabrication facility, and (2) MOX fuel from the latter to a reactor. These transport modes and differing forms of plutonium are considered in terms of: their proliferation resistance and safeguards; environmental and safety aspects; and economic aspects. It is tentatively proposed that the transport of plutonium could continue by air or sea where long distances are involved and by road or rail over shorter distances; this would be acceptable from the non-proliferation, environmental impact and economic aspects - there may be advantages in protection if plutonium is transported in the form of mixed oxide

  14. Recovery of americium-241 from aged plutonium metal

    International Nuclear Information System (INIS)

    Gray, L.W.; Burney, G.A.; Reilly, T.A.; Wilson, T.W.; McKibben, J.M.

    1980-12-01

    About 5 kg of ingrown 241 Am was recovered from 850 kg of aged plutonium using a process developed specifically for Savannah River Plant application. The aged plutonium metal was first dissolved in sulfamic acid. Sodium nitrite was added to oxidize the plutonium to Pu(IV) and the residual sulfamate ion was oxidized to nitrogen gas and sulfate. The plutonium and americium were separated by one cycle of solvent extraction. The recovered products were subsequently purified by cation exchange chromatography, precipitated as oxalates, and calcined to the oxides. Plutonium processng was routine. Before cation exchange purification, the aqueous americium solution from solvent extraction was concentrated and stripped of nitric acid. More than 98% of the 241 Am was then recovered from the cation exchange column where it was effectively decontaminated from all major impurities except nickel and chromium. This partially purified product solution was concentrated further by evaporation and then denitrated by reaction with formic acid. Individual batches of americium oxalate were then precipitated, filtered, washed, and calcined. About 98.5% of the americium was recovered. The final product purity averaged 98% 241 AmO 2 ; residual impurities were primarily lead and nickel

  15. The flashcal process for the fabrication of fuel-metal oxides using the whiteshell roto-spray calciner

    International Nuclear Information System (INIS)

    Sridhar, T.S.

    1988-01-01

    A one-step, continuous, thermochemical calcination process, called the FLASHCAL (Flash Calcination) process has been developed for the production of single- and mixed-oxide powders of fuel metals (uranium, thorium and plutonium) from the respective nitrate solutions using the Whiteshell Roto-Spray Calciner (RSC). The metal-nitrate feed solution, either by itself or mixed with a suitable chemical reactant or additive, is converted to its oxide powder in the RSC at temperatures between 300 and 600 0 C. Rapid denitration takes place in the calciner, yielding the metal-oxide powders while simultaneously destroying any excess chemical additive and reaction by-products. In the production of precursor oxide powders suitable for fuel fabrication, the FLASHCAL process has advantages over batch calcination and other processes that involve precipitation and filtration steps because fewer processing and handling operations are needed. Results obtained with thorium nitrate and uranium nitrate-thorium nitrate mixtures indicate that some measure of control over the size distribution and morphology of the oxide product powders is possible in this process with the proper selection of chemical additive, as well as the operating parameters of the calciner

  16. Corrosion-resistant powder-metallurgy stainless steel powders and compacts therefrom

    International Nuclear Information System (INIS)

    Klar, E.; Ro, D.H.; Whitman, C.I.

    1980-01-01

    Disclosed is a process for improving the corrosion resistance of a stainless steel powder or compact thereof wherein the powder is produced by atomizing a melt of metals in an oxidizing environment whereby the resulting stainless steel powder is surface-enriched in silicon oxides. The process comprises adding an effective proportion of modifier metal to the melt prior to the atomization, the modifier metal selected from the group consisting of tin, aluminum, lead, zinc, magnesium, rare earth metals and like metals capable of enrichment about the surface of the resulting atomized stainless steel powder and effective under reductive sintering conditions in the depletion of the silicon oxides about the surface; and sintering the resulting atomized powder or a compact thereof under reducing conditions, the sintered powder or compact thereof being depleted in the silicon oxides and the corrosion resistance of the powder or compact thereof being improved thereby

  17. Effect of fabrication conditions on the properties of indium tin oxide powders

    International Nuclear Information System (INIS)

    Xie Wei

    2008-01-01

    This paper reports that indium tin oxide (ITO) crystalline powders are prepared by coprecipitation method. Fabrication conditions mainly as sintering temperature and Sn doping content are correlated with the phase, microstructure, infrared emissivity in and powder resistivity of indium tin oxides by means of x-ray diffraction, Fourier transform infrared, and transmission electron microscope. The optimum sintering temperature of 1350°C and Sn doping content 6∼8wt% are determined. The application of ITO in the military camouflage field is proposed. (condensed matter: electronic structure, electrical, magnetic, and optical properties)

  18. New automated pellet/powder assay system

    International Nuclear Information System (INIS)

    Olsen, R.N.

    1975-01-01

    This paper discusses an automated, high precision, pellet/ powder assay system. The system is an active assay system using a small isotopic neutron source and a coincidence detection system. The handling of the pellet powder samples has been automated and a programmable calculator has been integrated into the system to provide control and data analysis. The versatile system can assay uranium or plutonium in either active or passive modes

  19. Pulmonary lesions induced by inhaled plutonium in beagles

    International Nuclear Information System (INIS)

    Dagle, G.E.; Lund, J.E.; Park, J.F.

    1975-01-01

    The histopathologic features of pulmonary fibrosis and bronchiolo-alveolar carcinoma in beagles exposed to aerosols of plutonium oxide were reviewed. A hypothesis of the pathogenesis of radiation pneumonitis induced by inhalation of plutonium oxide was presented; this hypothesis included phagocytosis of plutonium particles, fibrosis responding to the necrosis, and alveolar cell hyperplasia compensating for alveolar cells killed by alpha radiation. Histopathologic features of the epithelial changes suggest a progression from hyperplasia to metaplasia and, finally, to bronchiolo-alveolar carcinoma. The possibility of concurrent radiation-induced lymphopenia contributing to the development of bronchiolo-alveolar carcinoma through a loss of immunologic surveillance was discussed

  20. Specification analysis of plutonium fuels : a potentiometric method for the determination of plutonium

    International Nuclear Information System (INIS)

    Vaidyanathan, S.; Natarajan, P.R.

    1977-01-01

    A potentiometric method for the routine determination of plutonium in the specification analysis of plutonium fuels is described. Plutonium is oxidized to Pu(VI) with AgO and Pu(VI) is reduced with Fe(II) after the destruction of excess AgO with sulphamic acid. The excess Fe(II) is titrated potentiometrically against K 2 Cr 2 O 7 , the titration being carried out by adding a concentrated titrant solution from a weight burette and a suitably diluted solution from another weight burette near the end. The overall relative standard deviation obtained in 326 analyses of a working standard solution by eight experimenters is 0.14 percent. (author)

  1. Surveillance of sealed containers with plutonium oxide materials (ms163)

    Science.gov (United States)

    Worl, Laura; Berg, John; Ford, Doris; Martinez, Max; McFarlan, Jim; Morris, John; Padilla, Dennis; Rau, Karen; Smith, Coleman; Veirs, Kirk; Hill, Dallas; Prenger, Coyne

    2000-07-01

    DOE is embarking upon a program to store large quantities of plutonium-bearing materials for up to fifty years. Materials destined for long-term storage include metals and oxides that are stabilized and packaged according to the DOE storage standard, where the packaging consists of two nested, welded, stainless steel containers. We have designed instrumented storage containers that mimic the inner storage can specified in the 3013 standard at both full- and small-scale capacities (2.4 liter and 0.005 liter, respectively), Figures 1 and 2. The containers are designed to maintain the volume to material mass ratio while allowing the gas composition and pressure to be monitored over time.

  2. 75 FR 41850 - Amended Notice of Intent to Modify the Scope of the Surplus Plutonium Disposition Supplemental...

    Science.gov (United States)

    2010-07-19

    ... and packaging capabilities, including direct metal oxidation, to fulfill plutonium storage..., disassemble nuclear weapons pits (a weapons component) and convert the plutonium metal to an oxide form for fabrication into mixed uranium-plutonium oxide (MOX) reactor fuel in the Mixed Oxide Fuel Fabrication Facility...

  3. Partial equilibrium in induced redox reactions of plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Nikol' skii, B P; Posvol' skii, M V; Krylov, L I; Morozova, Z P

    1975-01-01

    A study was made of oxidation-reduction reactions of Pu in buffer solutions containing bichromate and a reducing agent which reacted with hexavalent chromium at pH=3.5. In most cases sodium nitrite was used. A rather slow reduction of Pu (6) with NaNO/sub 2/ in the course of which tetravalent plutonium was formed via disproportionation reaction of plutonium (5), became very rapid upon the addition of bichromate to the solution. The yield of tetravalent plutonium increased with an increase in the concentration of NaNO/sub 2/ and the bichromate but never reached 100%. This was due to a simultaneous occurrenc of the induced oxidation reaction of Pu(4), leading to a partial equilibrium between the valence forms of plutonium in the nitrite-bichromate system which on the whole was in a nonequilibrium state. It was shown that in the series of reactions leading to the reduction of plutonium the presence of bivalent chromium was a necessary link.

  4. Gas atomized precursor alloy powder for oxide dispersion strengthened ferritic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Rieken, Joel [Iowa State Univ., Ames, IA (United States)

    2011-12-13

    Gas atomization reaction synthesis (GARS) was employed as a simplified method for producing precursor powders for oxide dispersion strengthened (ODS) ferritic stainless steels (e.g., Fe-Cr-Y-(Ti,Hf)-O), departing from the conventional mechanical alloying (MA) process. During GARS processing a reactive atomization gas (i.e., Ar-O2) was used to oxidize the powder surfaces during primary break-up and rapid solidification of the molten alloy. This resulted in envelopment of the powders by an ultra-thin (t < 150 nm) metastable Cr-enriched oxide layer that was used as a vehicle for solid-state transport of O into the consolidated microstructure. In an attempt to better understand the kinetics of this GARS reaction, theoretical cooling curves for the atomized droplets were calculated and used to establish an oxidation model for this process. Subsequent elevated temperature heat treatments, which were derived from Rhines pack measurements using an internal oxidation model, were used to promote thermodynamically driven O exchange reactions between trapped films of the initial Cr-enriched surface oxide and internal Y-enriched intermetallic precipitates. This novel microstructural evolution process resulted in the successful formation of nano-metric Y-enriched dispersoids, as confirmed using high energy X-ray diffraction and transmission electron microscopy (TEM), equivalent to conventional ODS alloys from MA powders. The thermal stability of these Y-enriched dispersoids was evaluated using high temperature (1200°C) annealing treatments ranging from 2.5 to 1,000 hrs of exposure. In a further departure from current ODS practice, replacing Ti with additions of Hf appeared to improve the Y-enriched dispersoid thermal stability by means of crystal structure modification. Additionally, the spatial distribution of the dispersoids was found to depend strongly on the original rapidly solidified microstructure. To exploit this, ODS microstructures were engineered from

  5. Spectrophotometers for plutonium monitoring in HB-line

    Energy Technology Data Exchange (ETDEWEB)

    Lascola, R. J. [Savannah River Site (SRS), Aiken, SC (United States); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States); Kyser, E. A. [Savannah River Site (SRS), Aiken, SC (United States); Immel, D. M. [Savannah River Site (SRS), Aiken, SC (United States); Plummer, J. R. [Savannah River Site (SRS), Aiken, SC (United States); Evans, E. V. [Savannah River Site (SRS), Aiken, SC (United States)

    2016-02-12

    This report describes the equipment, control software, calibrations for total plutonium and plutonium oxidation state, and qualification studies for the instrument. It also provides a detailed description of the uncertainty analysis, which includes source terms associated with plutonium calibration standards, instrument drift, and inter-instrument variability. Also included are work instructions for instrument, flow cell, and optical fiber setup, work instructions for routine maintenance, and drawings and schematic diagrams.

  6. An ab initio study of plutonium oxides surfaces; Etude ab initio des surfaces d'oxydes de Pu

    Energy Technology Data Exchange (ETDEWEB)

    Jomard, G.; Bottin, F.; Amadon, B

    2007-07-01

    By means of first-principles calculations, we have studied the atomic structure as well as the thermodynamic stability of various plutonium dioxide surfaces in function of their environment (in terms of oxygen partial pressure and temperature). All these simulations have been performed with the ABINIT code. It is well known that DFT fails to describe correctly plutonium-based materials since 5f electrons in such systems are strongly correlated. In order to go beyond DFT, we have treated PuO{sub 2} and {beta}-Pu{sub 2}O{sub 3} in a DFT+U framework. We show that the couple of parameters (U,J) that works well for pure Pu is also well designed for describing ground state (GS) properties of these two oxides. The major improvement with respect with DFT is that we are able to predict an insulating GS in agreement with experiments. The presence of a gap in the DOS (Density of States) of plutonium oxides should play a significant role in the predicted surface reactivity. However, performing DFT+U calculations on surfaces of plutonium oxide from scratch was too ambitious. That is why we decided, as a first step, to study the stability of the (100), (110) and (111) surfaces of PuO{sub 2} in a DFT-GGA framework. For each of these orientations, we considered various terminations. These ab initio results have been introduced in a thermodynamic model which allows us to predict the relative stability of the different terminations as a function of temperature and oxygen partial pressure (p{sub O{sub 2}}). We conclude that at room temperature and for p{sub O{sub 2}}{approx}10 atm., the polar O{sub 2}-(100) termination is favoured. The stabilization of such a polar stoichiometric surface is surprising and should be confirmed by DFT+U calculations before any final conclusion. (authors)

  7. Improved plutonium consumption in a pressurised water reactor

    International Nuclear Information System (INIS)

    Puill, A.; Bergeron, J.

    1995-01-01

    Our goal is to improve plutonium consumption in a dedicated PWR while limiting the production of minor actinides. For lack of proving the system's reliability, we stay in reasonable configurations in which power capacity is maintained. Three ways are investigated in determining the fuel assembly: (a) standard geometry with mixed oxide in enriched uranium base; (b) standard geometry with plutonium oxide included in an inert matrix; (c) new geometry with special all-plutonium consumption varies from 50 kg/TWeh (a) up to 140 kg/TWeh (b) (upper point). The new geometry with special all plutonium rods mixed with standard uranium rods appears promising with a burning rate of 92 kg/TWeh for a production of minor actinides of 10 kg/TWeh. (authors). 3 refs., 3 figs., 4 tabs

  8. Uranium-plutonium fuel for fast reactors

    International Nuclear Information System (INIS)

    Antipov, S.A.; Astafiev, V.A.; Clouchenkov, A.E.; Gustchin, K.I.; Menshikova, T.S.

    1996-01-01

    Technology was established for fabrication of MOX fuel pellets from co-precipitated and mechanically blended mixed oxides. Both processes ensure the homogeneous structure of pellets readily dissolvable in nitric acid upon reprocessing. In order to increase the plutonium charge in a reactor-burner a process was tested for producing MOX fuel with higher content of plutonium and an inert diluent. It was shown that it is feasible to produce fuel having homogeneous structure and the content of plutonium up to 45% mass

  9. Decorporation approach following rat lung contamination with a moderately soluble compound of plutonium using local and systemic Ca-DTPA combined chelation

    International Nuclear Information System (INIS)

    Gremy, Olivier; Bruel, Sylvie; Renault, Daniel; Van der Meeren, Anne; Tsapis, Nicolas

    2012-01-01

    Decorporation efficacy of prompt pulmonary delivery of DTPA dry powder was assessed following lung contamination with plutonium nitrate and compared to an intravenous injection of DTPA solution and a combined administration of both DTPA compounds. In addition, efficacy of a delayed treatment was assessed. In case of either early or late administration, insufflated DTPA was more efficient than intravenously injected DTPA in reducing the plutonium lung burden due to its high local concentration. Prompt treatment with DTPA powder was also more effective in limiting extrapulmonary deposits by removing the early transportable fraction of plutonium from lungs prior its absorption into blood. Translocation of DTPA from lungs to blood may also contribute to the decrease in extrapulmonary retention, as shown by reduced liver deposit after delayed pulmonary administration of DTPA. Efficacy of DTPA dry powder was further increased by the combined intravenous administration of DTPA solution for reducing extrapulmonary deposits of plutonium and promoting its urinary excretion. According to our results, the most effective treatment protocol for plutonium decorporation was the early pulmonary delivery of DTPA powder supplemented by an intravenous injection of DTPA solution. Following inhalation of plutonium as nitrate chemical form, this combined chelation therapy should provide a more effective method of treatment than conventional intravenous injection alone. At later stages following lung contamination, pulmonary administration of DTPA should also be considered as the treatment of choice for decreasing the lung burden. (authors)

  10. The first milligrams of plutonium

    International Nuclear Information System (INIS)

    Goldschmidt, B.

    1996-01-01

    This paper relates the discovery of the different plutonium chemical extraction processes in their historical context. The first experiments started during the second world war in 1942 with the American ''Metallurgical Laboratory'' project which brought together Arthur Compton, Enrico Fermi and Glenn Seaborg. During the same period, a competitive English-Canadian project, the ''Montreal Project'', was carried out to test different plutonium solvent extraction techniques. The author participated in both projects and joined the CEA in 1946, where he was in charge of the uranium and plutonium chemistry. By the end of 1949, his team could isolate the first milligrams of French plutonium from uranium oxide pellets of the ZOE reactor. In the beginning of 1952 he developed with his team the PUREX process. (J.S.)

  11. Chloride removal from plutonium alloy

    International Nuclear Information System (INIS)

    Holcomb, H.P.

    1983-01-01

    SRP is evaluating a program to recover plutonium from a metallic alloy that will contain chloride salt impurities. Removal of chloride to sufficiently low levels to prevent damaging corrosion to canyon equipment is feasible as a head-end step following dissolution. Silver nitrate and mercurous nitrate were each successfully used in laboratory tests to remove chloride from simulated alloy dissolver solution containing plutonium. Levels less than 10 ppM chloride were achieved in the supernates over the precipitated and centrifuged insoluble salts. Also, less than 0.05% loss of plutonium in the +3, +4, or +6 oxidation states was incurred via precipitate carrying. These results provide impetus for further study and development of a plant-scale process to recover plutonium from metal alloy at SRP

  12. Interstitial pressure dependence of the thermal conductivity of some rare earth oxide powders

    International Nuclear Information System (INIS)

    Pradeep, P.

    1997-01-01

    Thermal transport properties of powdered materials depend upon interstitial gas pressure. The present study reports the experimental results for the effective thermal conductivity of three rare earth oxide powders viz. yttrium oxide, samarium oxide, and gadolinium oxide, at various interstitial pressures by using transient plane source (TPS) method. A theoretical model is also proposed for the interpretation of the variation of the effective thermal conductivity with interstitial gas pressure. Its validity is found to be good in low pressure range of 45 mm Hg to normal pressure when compared with the experimental results. Also an attempt has been made to calculate the variation of thermal conductivity with interstitial pressure in the high pressure range up to 2 kbar using the proposed model. (author)

  13. The plutonium challenge for the future

    International Nuclear Information System (INIS)

    Gray, L.W.

    2000-01-01

    In this paper author deal with the weapons-usable plutonium and with the possibilities of their managing. Russia has not disclosed the amount of plutonium produced, but various estimates indicate that the production was about 130 tonnes. Production has been curtailed in Russia; three dual-purpose reactors still produce weapons-grade plutonium - two at Tomsk-7 (renamed Seversk) and one at Krasnoyarsk-26 (renamed Zheleznogorsk Mining and Chemical Combine). In a 1994 United States-Russian agreement that has yet to enter into force, Russia agreed to close the remaining operating reactors by the year 2000. Treaties between the United States and Russia have already cut the number of nuclear warheads from more than 10,000 to about 6,000 under START 1, which has been ratified, and to about 3,500 under START 2, which still awaits approval. If Russia and the United States conclude START 3, that number could drop to between 2,000 and 2,500. On September 2, 1998, the Presidents of the United States and Russia signed the 'Joint statement of principles for Management and Disposition of Plutonium, Designated as No Longer Required for Defense Purposes.' In this joint statement the Presidents affirm the intention of each country to remove by stages approximately 50 metric tons of plutonium and to convert the nuclear weapons programs, and to convert this material so that it can never be used in nuclear weapons. These 100 tonne of plutonium must be managed in proper way such that it becomes neither a proliferation for an environmental risk. The United States has proposed that it manage it's 50 tonnes by a dual approach-once through MOX burning of a portion of the plutonium and immobilization in a ceramic matrix followed by en- casement in high level waste glass. Russia has proposed that it manage its full 50 tonnes by burning in a reactor. The MOX program in the United States would bum the cleaner plutonium metal and residues. Weapons components would be converted to plutonium oxide

  14. PFPF canister counter for foreign plutonium (PCAS-3) hardware operations and procedures manual

    International Nuclear Information System (INIS)

    Menlove, H.O.; Baca, J.; Kroncke, K.E.; Miller, M.C.; Takahashi, S.; Seki, S.; Inose, S.; Yamamoto, T.

    1993-01-01

    A neutron coincidence counter has been designed for the measurement of plutonium powder contained in tall storage canisters. The counter was designed for installation in the Plutonium Fuel Production Facility fabrication plant. Each canister contains from one to five cans of PuO 2 . The neutron counter measures the spontaneous-fission rate from the plutonium and, when this is combined with the plutonium isotopic ratios, the plutonium mass is determined. The system can accommodate plutonium loadings up to 12 kg, with 10 kg being a typical loading. Software has been developed to permit the continuous operation of the system in an unattended mode. Authentication techniques have been developed for the system. This manual describes the system and its operation and gives performance and calibration parameters for typical applications

  15. Radiation damage and annealing in plutonium tetrafluoride

    Science.gov (United States)

    McCoy, Kaylyn; Casella, Amanda; Sinkov, Sergey; Sweet, Lucas; McNamara, Bruce; Delegard, Calvin; Jevremovic, Tatjana

    2017-12-01

    A sample of plutonium tetrafluoride that was separated prior to 1966 at the Hanford Site in Washington State was analyzed at the Pacific Northwest National Laboratory (PNNL) in 2015 and 2016. The plutonium tetrafluoride, as received, was an unusual color and considering the age of the plutonium, there were questions about the condition of the material. These questions had to be answered in order to determine the suitability of the material for future use or long-term storage. Therefore, thermogravimetric/differential thermal analysis and X-ray diffraction evaluations were conducted to determine the plutonium's crystal structure, oxide content, and moisture content; these analyses reported that the plutonium was predominately amorphous and tetrafluoride, with an oxide content near ten percent. Freshly fluorinated plutonium tetrafluoride is known to be monoclinic. During the initial thermogravimetric/differential thermal analyses, it was discovered that an exothermic event occurred within the material near 414 °C. X-ray diffraction analyses were conducted on the annealed tetrafluoride. The X-ray diffraction analyses indicated that some degree of recrystallization occurred in conjunction with the 414 °C event. The following commentary describes the series of thermogravimetric/differential thermal and X-ray diffraction analyses that were conducted as part of this investigation at PNNL.

  16. Radiation damage and annealing in plutonium tetrafluoride

    International Nuclear Information System (INIS)

    McCoy, Kaylyn; Casella, Amanda; Sinkov, Sergey

    2017-01-01

    A sample of plutonium tetrafluoride that was separated prior to 1966 at the Hanford Site in Washington State was analyzed at the Pacific Northwest National Laboratory (PNNL) in 2015 and 2016. The plutonium tetrafluoride, as received, was an unusual color and considering the age of the plutonium, there were questions about the condition of the material. These questions had to be answered in order to determine the suitability of the material for future use or long-term storage. Therefore, thermogravimetric/differential thermal analysis and X-ray diffraction evaluations were conducted to determine the plutonium's crystal structure, oxide content, and moisture content; these analyses reported that the plutonium was predominately amorphous and tetrafluoride, with an oxide content near ten percent. Freshly fluorinated plutonium tetrafluoride is known to be monoclinic. And during the initial thermogravimetric/differential thermal analyses, it was discovered that an exothermic event occurred within the material near 414 °C. X-ray diffraction analyses were conducted on the annealed tetrafluoride. The X-ray diffraction analyses indicated that some degree of recrystallization occurred in conjunction with the 414 °C event. This commentary describes the series of thermogravimetric/differential thermal and X-ray diffraction analyses that were conducted as part of this investigation at PNNL.

  17. Plutonium use in foreign countries (03)

    International Nuclear Information System (INIS)

    Otagaki, Takao

    2004-03-01

    European countries and Japan had been implementing the strategy of spent fuel reprocessing in order to use nuclear material to the maximum. Plutonium recovered from reprocessing, however, must be recycle on light water reactors (LWRs) because of considerable delay of fast reactor development. In Europe, much of experiences of plutonium recycling have been accumulated until now. Thus, the status of plutonium recycling up to the end of 2003 in France, Germany, The U.K., Belgium, Switzerland and other countries were studied based on the following scope. (1) Basic policy and present status of plutonium recycling in primary countries of France, Germany, The U.K., Belgium, Switzerland, and Sweden which plans to recycle a part of plutonium: Backend policy and the status of spent fuel management were studied, then integrated analysis and evaluation of the position of plutonium recycling in backend and the status of plutonium recycling development were performed. (2) Plan and experience of Mixed Oxide (MOX) fuel fabrication and reprocessing of spent fuels: The data and information on plan and experience of MOX fuel fabrication and reprocessing in foreign countries were collected. (3) Plutonium inventories: The data and information of plutonium inventories of foreign countries were collected. (author)

  18. Determination of the clean 4f peak shape in XPS for plutonium metal

    Energy Technology Data Exchange (ETDEWEB)

    Morrall, P. [AWE, Aldermaston, Reading, Berkshire RG7 4PR (United Kingdom)], E-mail: peter.morrall@awe.co.uk; Roussel, P. [AWE, Aldermaston, Reading, Berkshire RG7 4PR (United Kingdom); Jolly, L.; Brevet, A.; Delaunay, F. [Commissariat a l' Energie Atomique, Centre de Valduc, 21120 Is-sur-Tille (France)

    2009-03-15

    Many of the interesting properties observed with plutonium are ascribed to the influence of 5f electrons, and to the degree of localisation observed within these electrons. Indeed, changes in 5f localisation are sensitively reflected in the final states observed in core-level photoemission measurements. However, when analysing the 4f manifold of elemental plutonium, it is essential to obtain spectra without the influence of oxidation, which can easily be misinterpreted as 5f localisation. The ideal method to extract elemental plutonium 4f spectra is to remove any influence of oxidation from the 'clean' plutonium data by careful measurement of the oxygen 1s region, and the subsequent subtraction of the unwanted oxide features. However, in order to achieve this objective it is essential to determine the relative sensitivity factor (RSF) for plutonium 4f and the precise shape of the 4f features from plutonium sesqui-oxide. In this paper, we report an experimental determination of the RSF for the plutonium 4f manifold using experimental data captured from two different Vacuum Generators spectrometers; an ESCALAB Mk II and an ESCALAB 220i.

  19. Powder of a copper oxide superconductor precursor, fabrication process and use for the preparation of superconducting oxide

    International Nuclear Information System (INIS)

    Dehaudt, P.

    1990-01-01

    The precursor powder comprises at least a copper compound (hydroxide, oxide and hydroxynitrates), at least a rare earth and/or yttrium compound (nitrates, hydroxides and hydroxynitrates) or bismuth oxide and at least an alkaline earth nitrate. It can be prepared by atomization drying of a suspension a copper precipitate or coprecipitate and other elements of the superconducting oxide in solution [fr

  20. A rapid and specific titrimetric method for the precise determination of plutonium using redox indicator

    International Nuclear Information System (INIS)

    Chitnis, R.T.; Dubey, S.C.

    1976-01-01

    A simple and rapid method for the determination of plutonium in plutonium nitrate solution and its application to the purex process solutions is discussed. The method involves the oxidation of plutonium to Pu(VI) with the help of argentic oxide followed by the destruction of the excess argentic oxide by means of sulphamic acid. The determination of plutonium is completed by adding ferrous ammonium sulphate solution which reduces Pu(VI) to Pu(IV) and titrating the excess ferrous with standard potassium dichromate solution using sodium diphenylamine sulphonate as the internal indicator. The effect of the various reagents add during the oxidation and reduction of plutonium, on the final titration has been investigated. The method works satisfactorily for the analysis of plutonium in the range of 0.5 to 5 mg. The precision of the method is found to be within 0.1%. (author)

  1. Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

    International Nuclear Information System (INIS)

    Trellue, Holly R.

    2004-01-01

    The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

  2. ( Citrus maxima ) peel powder on lipid oxidation of

    African Journals Online (AJOL)

    The presence of phenolic compounds in Shaddock peel powder (SPP) and its effect on lipid oxidation in cooked and raw semimembranosus (SM) muscle in goats were examined. The SPP was applied to cooked and raw minced SM at the rate of 0, 0.5, 1.0 and 1.5% of the weight of the meat and compared with 0.01% ...

  3. Pneumatic transport system development: residuals and releases program at Westinghouse Cheswick site

    International Nuclear Information System (INIS)

    Larouere, P.J.; Shoulders, J.L.

    1979-01-01

    Plutonium oxide and uranium oxide powders are processed within glove boxes or within confinement systems during the fabrication of mixed oxide (MOX) pellets for recycle fuel. The release of these powders to the glove box or to the confinement results in some airborne material that is deposited in the enclosure or is carried in the air streams to the effluent air filtration system. Release tests on simulated leaks in pneumatic transport equipment and release tests on simulated failures with powder blending equipment were conducted. A task to develop pneumatic transport for the movement of powders within an MOX fabrication plant has been underway at the Westinghouse Research Laboratories. While testing and evaluating selected pneumatic transport components on a full scale were in progress, it was deemed necessary that final verification of the technology would have to be performed with plutonium-bearing powders because of the marked differences in certain properties of plutonium from those of uranium oxides. A smaller was designed and constructed for the planned installation in glove boxes at the Westinghouse Plutonium Fuel Development Laboratory. However, prior to use with plutonium it was agreed that this system be set up and tested with uranium oxide powder. The test program conducted at the Westinghouse Cheswick site was divided into two major parts. The first of these examined the residuals left as a result of the pneumatic transport of nuclear fuel powders and verified the operability of this one-third scale system. The second part of the program studied the amount of powder released to the air when off-standard process procedures or maintenance operations were conducted on the pneumatic transport system. Air samplers located within the walk-in box housing the transport loop were used to measure the solids concentration in the air. From this information, the total amount of airborne powder was determined

  4. Kinetics of tetravalent plutonium oxidation by cerium (4)

    Energy Technology Data Exchange (ETDEWEB)

    Nikitina, G P; Shumakov, V G; Egorova, V P

    1975-01-01

    Stoichiometry and kinetics of the Pu(4) + Ce(4) reaction is studied by spectrophotometric method at 5-30 deg C in nitric acid solutions (..mu..=(HNO/sub 3/)+(NaNO/sub 3/) = 1.0 - 5.7; (Ce(4)) = 5.10/sup -5/ - 1.2.10/sup -3/ g-ion/1, (Pu(4)) =1.10/sup -5/ - 8.5.10/sup -4/ g-ion/1. Oxidation of one Pu(4) ion to a hexavalent state requires two Ce(4) ions. The plutonium oxidation is not complicated by by-processes. Reverse Ce(3) + PuO/sub 2//sup +/ reaction does not contribute essentially to the process at (Ce(3)) 2.6.10/sup -2/ g-ion/1. The reaction rate obeys the kinetic equation - d(Pu(4))/dt = ksub(eff)(Pu(4))(Ce(4))/a sub(+-HNO/sub 3/). The thermodynamical activation parameters are found for solutions at the varied nitric acid concentrations:..delta..Gsup(not equal) = 17.0+-0.2 and ..delta..Hsup(not equal) = 21.8+-2 kcal/mole(..mu..=1.0); ..delta..Gsup(not equal) = 19.1+-0.1 and ..delta..Hsup(not equal) = 23.1+-1 kcal/mole(..mu..=4.9). The reaction mechanism is discussed in terms of the theory of absolute reaction rates and the model of long-range charge transfer.

  5. Purifications of calcium carbonate and molybdenum oxide powders for neutrinoless double beta decay experiment, AMoRE

    International Nuclear Information System (INIS)

    Park, HyangKyu

    2015-01-01

    The AMoRE (Advanced Mo based Rare process Experiment) collaboration is going to use calcium molybdate crystals to search for neutrinoless double beta decay of 100 Mo isotope. In order to make the crystal, we use calcium carbonate and molybdenum oxide powders as raw materials. Therefore it is highly necessary to reduce potential sources for radioactive backgrounds such as U and Th in the powders. In this talk, we will present our studies for purification of calcium carbonate and molybdenum oxide powders

  6. Phase Transformation Behavior of Oxide Particles Formed in Mechanically Alloyed Fe-5Y{sub 2}O{sub 3} Powder

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ga Eon; Choi, Jung-Sun; Noh, Sanghoon; Kang, Suk Hoon; Choi, Byoung Kwon; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Deajeon (Korea, Republic of); Kim, Young Do [Hanyang University, Seoul (Korea, Republic of)

    2017-05-15

    The phase transformation behavior of the oxides formed in mechanically alloyed Fe-5Y{sub 2}O{sub 3} powder is investigated. Non-stoichiometric Y-rich and Fe-rich oxides with sizes of less than 300 nm are observed in the mechanically alloyed powder. The diffusion and redistribution reactions of the elements in these oxides during heating of the powder above 800 ℃ were observed, and these reactions result in the formation of a Y{sub 3}Fe{sub 5}O{sub 12} phase after heating at 1050 ℃. Thus, it is considered that the Y{sub 2}O{sub 3} powder and some Fe powder are formed from the non-stoichiometric Y-rich and Fe-rich oxides after the mechanical alloying process, and a considerable energy accumulated during the mechanical alloying process leads to a phase transformation of the Y-rich and Fe-rich oxides to Y{sub α}Fe{sub β}O{sub γ}-type phase during heating.

  7. Separation of plutonium oxide nanoparticles and colloids

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Richard E.; Skanthakumar, S.; Soderholm, L. [Argonne National Laboratory, IL (United States). Chemical Sciences and Engineering Division

    2011-11-18

    Colloidal plutonium is an important component of Pu aqueous speciation. Pu colloids are problematic in nuclear separations and are a potential transport vector in the environment. Using a mixture of n-octanol and trichloroacetic acid a selective and reversible separation of these particles can be achieved by exploiting their surface reactivity. [German] Kolloidales Plutonium ist ein wichtiger Bestandteil in waessrigen Pu-Bereitungen. Pu-Kolloide sind problematisch bei der Wiederaufbereitung von Kernmaterial und bilden einen potenziellen Transportvektor in die Umwelt. Mit einem Loesungsmittelgemisch aus n-Octanol und Trichloressigsaeure gelingt die selektive und reversible Trennung dieser Partikel durch Ausnutzung ihrer Oberflaechenreaktivitaet.

  8. Plutonium vulnerability issues at Hanford's Plutonium Finishing Plant

    International Nuclear Information System (INIS)

    Feldt, E.; Templeton, D.W.; Tholen, E.

    1995-01-01

    The Plutonium Finishing Plant (PFP) at the Hanford, Washington Site was operated to produce plutonium (Pu) metal and oxide for national defense purposes. Due to the production requirements and methods utilized to meet national needs and the abrupt shutdown of the plant in the late 1980s, the plant was left in a condition that poses a risk of radiation exposure to plant workers, of accidental radioactive material release to the environment, and of radiation exposure to the public. In early 1994, an Environmental Impact Statement (EIS) to determine the best methods for cleaning out and stabilizing Pu materials in the PFP was started. While the EIS is being prepared, a number of immediate actions have been completed or are underway to significantly reduce the greatest hazards in the PFP. Recently, increased attention his been paid to Pu risks at Department of Energy (DOE) facilities resulting in the Department-wide Plutonium Vulnerability Assessment and a recommendation by the Defense Nuclear Facilities Safety Board (DNFSB) for DOE to develop integrated plans for managing its nuclear materials

  9. Plutonium uniqueness

    International Nuclear Information System (INIS)

    Silver, G.L.

    1984-01-01

    A standard is suggested against which the putative uniqueness of plutonium may be tested. It is common folklore that plutonium is unique among the chemical elements because its four common oxidation states can coexist in the same solution. Whether this putative uniqueness appears only during transit to equilibrium, or only at equilibrium, or all of the time, is not generally made clear. But while the folklore may contain some truth, it cannot be put to test until some measure of 'uniqueness' is agreed upon so that quantitative comparisons are possible. One way of measuring uniqueness is as the magnitude of the product of the mole fractions of the element at equilibrium. A 'coexistence index' is defined and discussed. (author)

  10. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO 2 powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO 2 powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule

  11. Low-level detection and quantification of Plutonium(III, IV, V,and VI) using a liquid core waveguide

    International Nuclear Information System (INIS)

    Wilson, Richard E.; Hu, Yung-Jin; Nitsche, Heino

    2003-01-01

    Understanding the aqueous chemistry of plutonium, in particular in environmental conditions, is often complicated by plutonium's complex redox chemistry. Because plutonium possesses four oxidation states, all of which can coexist in solution, a reliable method for the identification of these oxidation states is needed. The identification of plutonium oxidation states at low levels in aqueous solution is often accomplished through an indirect determination using series of liquid-liquid extraction procedures using oxidation state specific reagents such as HDEHP and TTA. While these methods, coupled with radioactive counting techniques provide superior limits of detection they may influence the plutonium redox equilibrium, are time consuming, waste intensive and costly. Other analytical methods such as mass spectrometry and radioactive counting as stand alone methods provide excellent detection limits but lack the ability to discriminate between the oxidation states of the plutonium ions in solution

  12. Decontaminaion of metals containing plutonium and americium

    International Nuclear Information System (INIS)

    Seitz, M.G.; Gerding, T.J.; Steindler, M.J.

    1979-06-01

    Melt-slagging (melt-refining) techniques were evaluated as a decontamination and consolidation step for metals contaminated with oxides of plutonium and americium. Experiments were performed in which mild steel, stainless steel, and nickel contaminated with oxides of plutonium and americium were melted in the presence of silicate slags of various compositions. The metal products were low in contamination, with the plutonium and americium strongly fractionated to the slags. Partition coefficients (plutonium in slag/plutonium in steel) of 7 x 10 6 were measured with boro-silicate slag and of 3 x 10 6 with calcium, magnesium silicate slag. Decontamination of metals containing as much as 14,000 ppM plutonium appears to be as efficient as for metals with plutonium levels of 400 ppM. Staged extraction, that is, a remelting of processed metal with clean slag, results in further decontamination of the metal. The second extraction is effective with either resistance-furnace melting or electric-arc melting. Slag adhering to the metal ingots and in defects within the ingots is in the important contributors to plutonium retained in processed metals. If these sources of plutonium are controlled, the melt-refining process can be used on a large scale to convert highly contaminated metals to homogeneous and compact forms with very low concentrations of plutonium and americium. A conceptual design of a melt-refining process to decontaminate plutonium- and americium-contaminated metals is described. The process includes single-stage refining of contaminated metals to produce a metal product which would have less than 10 nCi/g of TRU-element contamination. Two plant sizes were considered. The smaller conceptual plant processes 77 kg of metal per 8-h period and may be portable.The larger one processes 140 kg of metal per 8-h period, is stationary, and may be near te maximum size that is practical for a metal decontamination process

  13. Simulation and control synthesis for a pulse column separation system for plutonium--uranium recovery

    International Nuclear Information System (INIS)

    McCutcheon, E.B.

    1975-05-01

    Control of a plutonium-uranium partitioning column was studied using a mathematical model developed to simulate the dynamic response and to test postulated separation mechanisms. The column is part of a plutonium recycle flowsheet developed for the recovery of plutonium and uranium from metallurgical scrap. In the first step of the process, decontamination from impurities is achieved by coextracting plutonium and uranium in their higher oxidation states. In the second step, reduction of the plutonium to a lower oxidation state allows partitioning of the plutonium and uranium. The use of hydroxylamine for the plutonium reduction in this partitioning column is a unique feature of the process. The extraction operations are carried out in pulse columns. (U.S.)

  14. Plutonium immobilization plant using ceramic in existing facilities at the Savannah River site

    International Nuclear Information System (INIS)

    DiSabatino, A.

    1998-01-01

    The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources, and through a ceramic immobilization process converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. This immobilization process is shown conceptually in Figure 1-1. The objective is to make an immobilized form, suitable for geologic disposal, in which the plutonium is as inherently unattractive and inaccessible as the plutonium in spent fuel from commercial reactors. The ceramic immobilization alternative presented in this report consists of first converting the surplus material to an oxide, followed by incorporating the plutonium oxide into a titanate-based ceramic material that is placed in metal cans

  15. Standard specification for nuclear-grade zirconium oxide powder

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification defines the physical and chemical requirements for zirconium oxide powder intended for fabrication into shapes, either entirely or partially of zirconia, for use in a nuclear reactor core. 1.2 The material described herein shall be particulate in nature. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  16. Plutonium again (smuggling and movements)

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    A link is discounted between nuclear proliferation and the recently discovered smuggled plutonium from the former Soviet Union at Munich airport and other places in Germany. It is argued that governments wishing to obtain nuclear materials to develop a weapons programme would not arrange to have it smuggled in a suitcase. Instead, it is speculated that a link exists between the plutonium smuggling incidents and the desire to promote the production of mixed oxide (MOX) fuel. Such incidents, by further raising public anxiety, may be intended to turn public opinion in favour of MOX fuel production as a sensible way of getting rid of surplus plutonium. (UK)

  17. Plutonium use in foreign countries (01)

    International Nuclear Information System (INIS)

    Otagaki, Takao

    2002-03-01

    European countries and Japan had been implementing the strategy of spent fuel reprocessing in order to use nuclear material to the maximum. Plutonium recovered from reprocessing, however, must be recycle on light water reactors (LWRs) because of considerable delay of fast reactor development. In Europe, much of experience of plutonium recycling have been accumulated until now. Thus, the status of plutonium recycling up to the end of 2001 in France, Germany, The U.K., Belgium, Switzerland and other countries were studied based on the following scope. (1) Basic policy and present status of plutonium recycling in primary countries of France, Germany, The U.K., Belgium, Switzerland, and Sweden which recently appears the move of recycling a part of plutonium. Backend policy and the status of spent fuel management were studied, then integrated analysis and evaluation of the position of plutonium recycling in backend and the status of plutonium recycling development were performed. (2) Plan and experience of Mixed Oxide (MOX) fuel fabrication and reprocessing of spent fuels. The data and information on plan and experience of MOX fuel fabrication and reprocessing in foreign countries were collected. (3) Plutonium inventories. The data and information on plutonium inventories of foreign countries were collected. (author)

  18. Plutonium use in foreign countries (99)

    International Nuclear Information System (INIS)

    Otagaki, Takao

    2000-03-01

    European countries and Japan had been implementing the strategy of spent fuel reprocessing in order to use nuclear material to the maximum. Plutonium recovered from reprocessing, however, must be recycle on light water reactors (LWRs) because of considerable delay of fast reactor development. In Europe, much of experience of plutonium recycling have been accumulated until now. Thus, the status of plutonium recycling up to the end of 1999 in France, Germany, The U.K., Belgium, Switzerland and other countries were studied based on the following scope. (1) Basic policy and present status of plutonium recycling in primary countries of France, Germany, The U.K., Belgium, Switzerland, and Sweden which recently appears the move to recycling a part of plutonium backend policy and the status of spent fuel management were studied, then integrated analysis and evaluation of the position of plutonium recycling in backend and the status of plutonium recycling development were performed. (2) Plan and experience of Mixed Oxide (MOX) fuel fabrication and reprocessing of spent fuels. The data and information on plan and experience of MOX fuel fabrication and reprocessing in foreign countries were collected. (3) Plutonium inventories. The data and information on plutonium inventories of foreign counties were collected. (author)

  19. Plutonium use in foreign countries. (04)

    International Nuclear Information System (INIS)

    Otagaki, Takao

    2005-03-01

    European countries and Japan had been implementing the strategy of spent fuel reprocessing in order to use nuclear material to the maximum. Plutonium recovered from reprocessing, however, must be recycle on light water reactors (LWRs) because of considerable delay of fast reactor development. In Europe, much of experience of plutonium recycling have been accumulated until now. Thus, the status of plutonium recycling up to the end of 2004 in France, Germany, The U.K., Belgium, Switzerland and other countries were studied based on the following scope. (1) Basic policy and present status of plutonium recycling in primary countries of France, Germany, the U.K., Belgium, Switzerland, and Sweden which plans to recycle a limited amount of plutonium: Backend policy and the status of spent fuel management were studied, then integrated analysis and evaluation of the position of plutonium recycling in backend and the status of plutonium recycling development were performed. (2) Plan and experience of Mixed Oxide (MOX) fuel fabrication and reprocessing of spent fuels: The data and information on plan and experience of MOX fuel fabrication and reprocessing in foreign countries were collected. (3) Plutonium inventories: The data and information on plutonium inventories of foreign countries were collected. (author)

  20. Plutonium waste incineration using pyrohydrolysis

    International Nuclear Information System (INIS)

    Meyer, M.L.

    1991-01-01

    Waste generated by Savannah River Site (SRS) plutonium operations includes a contaminated organic waste stream. A conventional method for disposing of the organic waste stream and recovering the nuclear material is by incineration. When the organic material is burned, the plutonium remains in the incinerator ash. Plutonium recovery from incinerator ash is highly dependent on the maximum temperature to which the oxide is exposed. Recovery via acid leaching is reduced for a high fired ash (>800 degree C), while plutonium oxides fired at lower decomposition temperatures (400--800 degrees C) are more soluble at any given acid concentration. To determine the feasibility of using a lower temperature process, tests were conducted using an electrically heated, controlled-air incinerator. Nine nonradioactive, solid, waste materials were batch-fed and processed in a top-heated cylindrical furnace. Waste material processing was completed using a 19-liter batch over a nominal 8-hour cycle. A processing cycle consisted of 1 hour for heating, 4 hours for reacting, and 3 hours for chamber cooling. The water gas shift reaction was used to hydrolyze waste materials in an atmosphere of 336% steam and 4.4% oxygen. Throughput ranged from 0.14 to 0.27 kg/hr depending on the variability in the waste material composition and density

  1. Chemistry of plutonium revealed

    International Nuclear Information System (INIS)

    Connick, R.E.

    1992-01-01

    In 1941 one goal of the Manhattan Project was to unravel the chemistry of the synthetic element plutonium as rapidly as possible. In this paper the work carried out at Berkeley from the spring of 1942 to the summer of 1945 is described briefly. The aqueous chemistry of plutonium is quite remarkable. Important insights were obtained from tracer experiments, but the full complexity was not revealed until macroscopic amounts (milligrams) became available. Because processes for separation from fission products were based on aqueous solutions, such solution chemistry was emphasized, particularly precipitation and oxidation-reduction behavior. The latter turned out to be unusually intricate when it was discovered that two more oxidation states existed in aqueous solution than had previously been suspected. Further, an equilibrium was rapidly established among the four aqueous oxidation states, while at the same time any three were not in equilibrium. These and other observations made while doing a crash study of a previously unknown element are reported

  2. Accountability methods for plutonium and uranium: the NRC manuals

    Energy Technology Data Exchange (ETDEWEB)

    Gutmacher, R.G.; Stephens, F.B.

    1977-09-28

    Four manuals containing methods for the accountability of plutonium nitrate solutions, plutonium dioxide, uranium dioxide and mixed uranium-plutonium oxide have been prepared by us and issued by the U.S. Nuclear Regulatory Commission. A similar manual on methods for the accountability of uranium and plutonium in reprocessing plant dissolver solutions is now in preparation. In the present paper, we discuss the contents of the previously issued manuals and give a preview of the manual now being prepared.

  3. Accountability methods for plutonium and uranium: the NRC manuals

    International Nuclear Information System (INIS)

    Gutmacher, R.G.; Stephens, F.B.

    1977-01-01

    Four manuals containing methods for the accountability of plutonium nitrate solutions, plutonium dioxide, uranium dioxide and mixed uranium-plutonium oxide have been prepared by us and issued by the U.S. Nuclear Regulatory Commission. A similar manual on methods for the accountability of uranium and plutonium in reprocessing plant dissolver solutions is now in preparation. In the present paper, we discuss the contents of the previously issued manuals and give a preview of the manual now being prepared

  4. The US plutonium materials conversion program in Russia

    International Nuclear Information System (INIS)

    Zygmunt, S.J.; Mason, C.F.V.; Hahn, W.K.

    2000-01-01

    Progress has been made in Russia towards the conversion of weapons-grade plutonium (w-Pu) into plutonium oxide (PuO 2 ) suitable for further manufacture into mixed oxide (MOX) fuels. This program was started in 1998 in response to US proliferation concerns and the acknowledged international need to decrease the available weapons-grade Pu. A similar agenda is being followed in the US to address disposition of US surplus weapons-grade Pu. In Russia a conversion process has been selected and a site proposed. This paper discusses the present state of the program in support of this future operating facility that will process up to 5 metric tons of plutonium a year. (authors)

  5. Safety problems relating to plutonium recycling in light water reactors

    International Nuclear Information System (INIS)

    Devillers, C.; Frison, J.M.; Mercier, J.P.; Revais, J.P.

    1991-01-01

    This paper describes the specific nature, as regards safety, of the mixed oxide (MOX) fuel cycle, with the exception of safety problems relating to the operation of nuclear power plants. These specific characteristics are due mainly to the presence of plutonium in fresh fuel and to the higher plutonium and transuranic element content in spent fuel assemblies. The fuel cycle steps analysed here are the transport of plutonium oxide, the manufacture of MOX fuel assemblies, the transport of fresh and spent fuel assemblies and the processing of spent fuel assemblies

  6. Safety problems relating to plutonium recycling in light water reactors

    International Nuclear Information System (INIS)

    Devillers, C.; Frison, J.M.; Mercier, J.P.; Revais, J.P

    1991-01-01

    This paper describes the specific nature, as regards safety, of the mixed oxide (MOX) fuel cycle, with the exception of safety problems relating to the operation of nuclear power plants. These specific characteristics are due mainly to the presence of plutonium in fresh fuel and to the higher plutonium and transuranic element content in spent fuel assemblies. The fuel cycle steps analysed here are the transport of plutonium oxide, the manufacture of MOX fuel assemblies, the transport of fresh and spent fuel assemblies and the processing of spent fuel assemblies. (author) [fr

  7. Phenomenology of uranium-plutonium homogenization in nuclear fuels

    International Nuclear Information System (INIS)

    Marin, J.M.

    1988-01-01

    The uranium and plutonium cations distribution in mixed oxide fuels (U 1-y Pu y )O 2 with y ≤ 0.1 has been studied in laboratory with industrial fabrication methods. Our experiences has showed a slow cations migration. In the substoichiometry (UPu)O 2-x the diffusion is in connection with the plutonium valence which is an indicator of the oxidoreduction state of the crystal lattice. The plutonium valence is in connection with the oxygen ion deficit in order to compensate the electrical charge. The oxygen ratio of the solid depends of the oxygen partial pressure prevailing at the time of product elaboration but it can be modified by impurities. These impurities permit to increase or decrease the fuel characteristics and performances. An homogeneity analysis methodology is proposed, its objective is to classify the mixed oxide fuels according to the uranium and plutonium ions distribution [fr

  8. Verification station for Sandia/Rockwell Plutonium Protection system

    International Nuclear Information System (INIS)

    Nicholson, N.; Hastings, R.D.; Henry, C.N.; Millegan, D.R.

    1979-04-01

    A verification station has been designed to confirm the presence of plutonium within a container module. These container modules [about 13 cm (5 in.) in diameter and 23 cm (9 in.) high] hold sealed food-pack cans containing either plutonium oxide or metal and were designed by Sandia Laboratories to provide security and continuous surveillance and safety. After the plutonium is placed in the container module, it is closed with a solder seal. The verification station discussed here is used to confirm the presence of plutonium in the container module before it is placed in a carousel-type storage array inside the plutonium storage vault. This measurement represents the only technique that uses nuclear detectors in the plutonium protection system

  9. The chemistry of plutonium revealed

    International Nuclear Information System (INIS)

    Connick, R.E.

    1990-01-01

    In 1941 one goal of the Manhattan Project was to unravel the chemistry of the synthetic element plutonium as rapidly as possible. Important insights were obtained from tracer experiments, but the full complexity of plutonium chemistry was not revealed until macroscopic amounts (milligrams) became available. Because processes for separation from fission products were aqueous solution based, such solution chemistry was emphasized, particularly precipitation and oxidation-reduction behavior. The latter turned out to be unusually intricate when it was discovered that two more oxidation states existed in aqueous solution than had previously been suspected. Further, it was found that an equilibrium was rapidly established among the four aqueous oxidation states while at the same time any three were not in equilibrium. These and other observations made while doing a crash study of a previously unknown element will be reported

  10. Nonequilibrium method for estimating the first hydrolysis constant of tetravalent plutonium

    International Nuclear Information System (INIS)

    Silver, G.L.

    2010-01-01

    Six discrete formulas are used to estimate the equilibrium constant of the first hydrolysis reaction of tetravalent plutonium. They apply the pH, the oxidation number, one equilibrium constant, and fractions of two of the plutonium oxidation states. The new formulas are not restricted to the equilibrium condition. (author)

  11. Aluminum powder metallurgy processing

    Energy Technology Data Exchange (ETDEWEB)

    Flumerfelt, J.F.

    1999-02-12

    The objective of this dissertation is to explore the hypothesis that there is a strong linkage between gas atomization processing conditions, as-atomized aluminum powder characteristics, and the consolidation methodology required to make components from aluminum powder. The hypothesis was tested with pure aluminum powders produced by commercial air atomization, commercial inert gas atomization, and gas atomization reaction synthesis (GARS). A comparison of the GARS aluminum powders with the commercial aluminum powders showed the former to exhibit superior powder characteristics. The powders were compared in terms of size and shape, bulk chemistry, surface oxide chemistry and structure, and oxide film thickness. Minimum explosive concentration measurements assessed the dependence of explosibility hazard on surface area, oxide film thickness, and gas atomization processing conditions. The GARS aluminum powders were exposed to different relative humidity levels, demonstrating the effect of atmospheric conditions on post-atomization processing conditions. The GARS aluminum powders were exposed to different relative humidity levels, demonstrating the effect of atmospheric conditions on post-atomization oxidation of aluminum powder. An Al-Ti-Y GARS alloy exposed in ambient air at different temperatures revealed the effect of reactive alloy elements on post-atomization powder oxidation. The pure aluminum powders were consolidated by two different routes, a conventional consolidation process for fabricating aerospace components with aluminum powder and a proposed alternative. The consolidation procedures were compared by evaluating the consolidated microstructures and the corresponding mechanical properties. A low temperature solid state sintering experiment demonstrated that tap densified GARS aluminum powders can form sintering necks between contacting powder particles, unlike the total resistance to sintering of commercial air atomization aluminum powder.

  12. Synthesis and characterization of nanocomposite powders of calcium phosphate/titanium oxide for biomedical applications

    Energy Technology Data Exchange (ETDEWEB)

    Delima, S.A.; Camargo, N.H.A.; Souza, J.C.P.; Gemelli, E., E-mail: sarahamindelima@hotmail.com, E-mail: dem2nhac@joinville.udesc.br, E-mail: souzajulio@joinville.udesc.br, E-mail: gemelli@joinville.udesc.br [Universidade do Estado de Santa Catarina (UDESC), Joinville, SC (Brazil). Centro de Ciencias Tecnologicas

    2009-07-01

    The nanostructured bioceramics of calcium phosphate are current themes of research and they are becoming important as bone matrix in regeneration of tissues in orthopedic and dental applications. Nanocomposite powders of calcium phosphate, reinforced with nanometric particles of titanium oxide, silica oxide and alumina oxid ealpha, are being widely studied because they offer new microstructures, nanostructures and interconnected microporosity with high superficial area of micropores that contribute to osteointegration and osteoinduction processes. This study is about the synthesis of nanocomposites powders of calcium phosphate reinforced with 1%, 2%, 3% and 5% in volume of titanium oxide and its characterization through the techniques of X-Ray Diffraction (XRD), Scanning Electron Microscopy (SEM), Differential Thermal Analysis (DTA), Thermogravimetry (TG) and Dilatometry. (author)

  13. Synthesis and characterization of nanocomposite powders of calcium phosphate/titanium oxide for biomedical applications

    International Nuclear Information System (INIS)

    Delima, S.A.; Camargo, N.H.A.; Souza, J.C.P.; Gemelli, E.

    2009-01-01

    The nanostructured bioceramics of calcium phosphate are current themes of research and they are becoming important as bone matrix in regeneration of tissues in orthopedic and dental applications. Nanocomposite powders of calcium phosphate, reinforced with nanometric particles of titanium oxide, silica oxide and alumina oxid ealpha, are being widely studied because they offer new microstructures, nanostructures and interconnected microporosity with high superficial area of micropores that contribute to osteointegration and osteoinduction processes. This study is about the synthesis of nanocomposites powders of calcium phosphate reinforced with 1%, 2%, 3% and 5% in volume of titanium oxide and its characterization through the techniques of X-Ray Diffraction (XRD), Scanning Electron Microscopy (SEM), Differential Thermal Analysis (DTA), Thermogravimetry (TG) and Dilatometry. (author)

  14. High resolution Transmission Electron Microscopy characterization of a milled oxide dispersion strengthened steel powder

    Energy Technology Data Exchange (ETDEWEB)

    Loyer-Prost, M., E-mail: marie.loyer-prost@cea.fr [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Merot, J.-S. [Laboratoire d’Etudes des Microstructures – UMR 104, CNRS/ONERA, BP72-29, Avenue de la Division Leclerc, 92 322, Châtillon (France); Ribis, J. [DEN-Service de Recherches de Métallurgie Appliquée, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Le Bouar, Y. [Laboratoire d’Etudes des Microstructures – UMR 104, CNRS/ONERA, BP72-29, Avenue de la Division Leclerc, 92 322, Châtillon (France); Chaffron, L. [DEN-Service de Recherches de Métallurgie Appliquée, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Legendre, F. [DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2016-10-15

    Oxide Dispersion Strengthened (ODS) steels are promising materials for generation IV fuel claddings as their dense nano-oxide dispersion provides good creep and irradiation resistance. Even if they have been studied for years, the formation mechanism of these nano-oxides is still unclear. Here we report for the first time a High Resolution Transmission Electron Microscopy and Energy Filtered Transmission Electron Microscopy characterization of an ODS milled powder. It provides clear evidence of the presence of small crystalline nanoclusters (NCs) enriched in titanium directly after milling. Small NCs (<5 nm) have a crystalline structure and seem partly coherent with the matrix. They have an interplanar spacing close to the (011) {sub bcc} iron structure. They coexist with larger crystalline spherical precipitates of 15–20 nm in size. Their crystalline structure may be metastable as they are not consistent with any Y-Ti-O or Ti-O structure. Such detailed observations in the as-milled grain powder confirm a mechanism of Y, Ti, O dissolution in the ferritic matrix followed by a NC precipitation during the mechanical alloying process of ODS materials. - Highlights: • We observed an ODS ball-milled powder by high resolution transmission microscopy. • The ODS ball-milled powder exhibits a lamellar microstructure. • Small crystalline nanoclusters were detected in the milled ODS powder. • The nanoclusters in the ODS milled powder are enriched in titanium. • Larger NCs of 15–20 nm in size are, at least, partly coherent with the matrix.

  15. Plutonium fires; Incendies de plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Mestre, E.

    1959-06-23

    The author reports an information survey on accidents which occurred when handling plutonium. He first addresses accidents reported in documents. He indicates the circumstances and consequences of these accidents (explosion in glove boxes, fires of plutonium chips, plutonium fire followed by filter destruction, explosion during plutonium chip dissolution followed by chip fire). He describes hazards associated with plutonium fires: atmosphere and surface contamination, criticality. The author gives some advices to avoid plutonium fires. These advices concern electric installations, the use of flammable solvents, general cautions associated with plutonium handling, venting and filtration. He finally describes how to fight plutonium fires, and measures to be taken after the fire (staff contamination control, atmosphere control)

  16. Standard specification for nuclear-grade beryllium oxide powder

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    This specification defines the physical and chemical requirements of nuclear-grade beryllium oxide (BeO) powder to be used in fabricating nuclear components. This specification does not include requirements for health and safety. It recognizes the material as a Class B poison and suggests that producers and users become thoroughly familiar with and comply to applicable federal, state and local regulations and handling guidelines. Special tests and procedures are given

  17. Disposing of the world's excess plutonium

    International Nuclear Information System (INIS)

    McCormick, J.M.; Bullen, D.B.

    1998-01-01

    The authors undertake three key objectives in addressing the issue of plutonium disposition at the end of the Cold War. First, the authors estimate the total global inventory of plutonium both from weapons dismantlement and civil nuclear power reactors. Second, they review past and current policy toward handling this metal by the US, Russia, and other key countries. Third, they evaluate the feasibility of several options (but especially the vitrification and mixed oxide fuel options announced by the Clinton administration) for disposing of the increasing amounts of plutonium available today. To undertake this analysis, the authors consider both the political and scientific problems confronting policymakers in dealing with this global plutonium issue. Interview data with political and technical officials in Washington and at the International Atomic Energy Agency in Vienna, Austria, and empirical inventory data on plutonium from a variety of sources form the basis of their analysis

  18. Procedure for plutonium determination using Pu(VI) spectra

    International Nuclear Information System (INIS)

    Walker, L.F.; Temer, D.J.; Jackson, D.D.

    1996-01-01

    This document describes a simple spectrophotometric method for determining total plutonium in nitric acid solutions based on the spectrum of Pu(VI). Plutonium samples in nitric acid are oxidized to Pu(VI) with Ce(IV) and the net absorbance at the 830 nm peak is measured

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

    Science.gov (United States)

    Mörsdorf, Alexander; Odnevall Wallinder, Inger; Hedberg, Yolanda

    2015-08-01

    The European chemical framework REACH requires that hazards and risks posed by chemicals, including alloys and metals, that are manufactured, imported or used in different products (substances or articles) are identified and proven safe for humans and the environment. Metals and alloys need hence to be investigated on their extent of released metals (bioaccessibility) in biologically relevant environments. Read-across from available studies may be used for similar materials. This study investigates the release of molybdenum and iron from powder particles of molybdenum metal (Mo), a ferromolybdenum alloy (FeMo), an iron metal powder (Fe), MoO2, and MoO3 in different synthetic body fluids of pH ranging from 1.5 to 7.4 and of different composition. Spectroscopic tools and cyclic voltammetry have been employed to characterize surface oxides, microscopy, light scattering and nitrogen absorption for particle characterization, and atomic absorption spectroscopy to quantify released amounts of metals. The release of molybdenum from the Mo powder generally increased with pH and was influenced by the fluid composition. The mixed iron and molybdenum surface oxide of the FeMo powder acted as a barrier both at acidic and weakly alkaline conditions. These findings underline the importance of the surface oxide characteristics for the bioaccessibility of metal alloys. Copyright © 2015 The Authors. Published by Elsevier Inc. All rights reserved.

  20. Plutonium working group report on environmental, safety and health vulnerabilities associated with the Department's plutonium storage. Volume I: Summary

    International Nuclear Information System (INIS)

    1994-11-01

    At the conclusion of the Cold War, the Department of Energy (DOE) stopped plutonium processing for nuclear weapons production. Facilities used for that purpose now hold significant quantities of plutonium in various forms. Unless properly stored and handled, plutonium can present environment, safety and health (ES ampersand H) hazards. Improperly stored plutonium poses a variety of hazards. When containers or packaging fail to fully protect plutonium metal from exposure to air, oxidation can occur and cause packaging failures and personnel contamination. Contamination can also result when plutonium solutions leak from bottles, tanks or piping. Plutonium in the form of scrap or residues generated by weapons production are often very corrosive, chemically reactive and difficult to contain. Buildings and equipment that are aging, poorly maintained or of obsolete design contribute to the overall problem. Inadvertent accumulations of plutonium of any form in sufficient quantities within facilities can result in nuclear criticality events that could emit large amounts of radiation locally. Contamination events and precursors of criticality events are causing safety and health concerns for workers at the Department's plutonium facilities. Contamination events also potentially threaten the public and the surrounding environment

  1. Plutonium transport for the oxide form. Air and road accidentology

    International Nuclear Information System (INIS)

    Degrange, J.P.; Hubert, P.; Pages, P.

    1989-07-01

    The aim of the study is to define the probability of accident that can be applied to a packaging for plutonium oxide transport. The analysis is designed essentially for packagings of few tons with impact resistance corresponding to a speed of 10 m/s but can be used for packagings with a resistance corresponding to IAEA tests. This work is based on analysis and processing of 2 accident files: one on road accidents, the other on air accidents. Computer codes are developed, from one hand, to obtain stress indicator distribution by Monte Carlo methods and on the other hand to link these distributions to itinerary parameters. Previous similar studies are examined and compared to figures obtained here and explained as far as possible [fr

  2. Radiation damage and annealing in plutonium tetrafluoride

    Energy Technology Data Exchange (ETDEWEB)

    McCoy, Kaylyn; Casella, Amanda; Sinkov, Sergey; Sweet, Lucas; McNamara, Bruce; Delegard, Calvin; Jevremovic, Tatjana

    2017-12-01

    Plutonium tetrafluoride that was separated prior to 1966 at the Hanford Site in Washington State was analyzed at the Pacific Northwest National Laboratory (PNNL) in 2015 and 2016. The plutonium tetrafluoride, as received, was an off-normal color and considering the age of the plutonium, there were questions about the condition of the material. These questions had to be answered in order to determine the suitability of the material for future use or long-term storage. Therefore, Thermogravimetric/Differential Thermal Analysis and X-ray Diffraction evaluations were conducted to determine the plutonium’s crystal structure, oxide content, and moisture content; these analyses reported that the plutonium was predominately amorphous and tetrafluoride, with an oxide content near ten percent. Freshly fluorinated plutonium tetrafluoride is known to be monoclinic. During the initial Thermogravimetric/Differential Thermal analyses, it was discovered that an exothermic event occurred within the material near 414°C. X-ray Diffraction analyses were conducted on the annealed tetrafluoride. The X-ray Diffraction analyses indicated that some degree of recrystallization occurred in conjunction with the 414°C event. The following commentary describes the series of Thermogravimetric/Differential Thermal and X-ray Diffraction analyses that were conducted as part of this investigation at PNNL, in collaboration with the University of Utah Nuclear Engineering Program.

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

    International Nuclear Information System (INIS)

    Silver, G.L.; Koenst, J.W.

    1977-01-01

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

  4. Development of Oxide Dispersion Strengthened (ODS) Ferritic Steel Through Powder Forging

    Science.gov (United States)

    Kumar, Deepak; Prakash, Ujjwal; Dabhade, Vikram V.; Laha, K.; Sakthivel, T.

    2017-04-01

    Oxide dispersion strengthened (ODS) ferritic steels are candidates for cladding tubes in fast breeder nuclear reactors. In this study, an 18%Cr ODS ferritic steel was prepared through powder forging route. Elemental powders with a nominal composition of Fe-18Cr-2 W-0.2Ti (composition in wt.%) with 0 and 0.35% yttria were prepared by mechanical alloying in a Simoloyer attritor under argon atmosphere. The alloyed powders were heated in a mild steel can to 1473 K under flowing hydrogen atmosphere. The can was then hot forged. Steps of sealing, degassing and evacuation are eliminated by using powder forging. Heating ODS powder in hydrogen atmosphere ensures good bonding between alloy powders. A dense ODS alloy with an attractive combination of strength and ductility was obtained after re-forging. On testing at 973 K, a loss in ductility was observed in yttria-containing alloy. The strength and ductility increased with increase in strain rate at 973 K. Reasons for this are discussed. The ODS alloy exhibited a recrystallized microstructure which is difficult to achieve by extrusion. No prior particle boundaries were observed after forging. The forged compacts exhibited isotropic mechanical properties. It is suggested that powder forging may offer several advantages over the traditional extrusion/HIP routes for fabrication of ODS alloys.

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

    International Nuclear Information System (INIS)

    Green, D.W.

    1980-03-01

    Measured and estimated spectroscopic data for thorium, uranium, and plutonium oxide vapor species have been used with the methods of statistical mechanics to calculate thermodynamic functions. Some inconsistencies between spectroscopic data and some thermodynamic data have been resolved by recalculating ΔH 0 /sub f/ (298.15 0 K) values for the vapor species of these oxides. Evaluation of the uncertainties in data, methods of estimating molecular parameters, and effects of assumptions have been discussed elsewhere. The tables of thermodynamic functions that were reported earlier have been revised principally because the low-frequency vibrational modes of UO 2 and UO 3 have now been measured. These new empirical data resulted in changes in the electronic contributions to the calculated thermodynamic functions of UO 2 and the estimated vibrational contributions for PuO 2 . In addition, some minor changes have been made in the methods of calculation of the electronic contributions for all molecules

  6. Plutonium determination by spectrophotometry of plutonium (VI): control of the nuclear fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Grison, J [Compagnie Generale des Matieres Nucleaires (COGEMA), Centre de la Hague, 50 - Cherbourg (France)

    1980-10-01

    The plutonium (VI) spectrophotometric determination, after AgO oxidation in 3 M nitric acid medium, is used for the running-control of the nuclear fuel reprocessing plant at La Hague. Analytical device used in glove-box or shielded-cell is briefly described. This method is fast, sensitive, unfailing and gives simple effluents. It is applied by day and night shifts, during Light Water Reactor fuel reprocessing campaign, for 0.5 mg/l up to 20 g/l plutonium solutions. Reference solution measurements have a 0.8 to 1.4 % relative standard deviation; duplicate plutonium determinations give a 0.3% relative standard deviation for sample analysis. There is a discrepancy (- 0.3% to - 0.9%) between the spectrophotometric method results and the isotopic dilution analysis.

  7. Plutonium determination by spectrophotometry of plutonium (VI): control of the nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Grison, J.

    1980-01-01

    The plutonium (VI) spectrophotometric determination, after AgO oxidation in 3 M nitric acid medium, is used for the running-control of the nuclear fuel reprocessing plant at La Hague. Analytical device used in glove-box or shielded-cell is briefly described. This method is fast, sensitive, unfailing and gives simple effluents. It is applied by day and night shifts, during Light Water Reactor fuel reprocessing campaign, for 0.5 mg/l up to 20 g/l plutonium solutions. Reference solution measurements have a 0.8 to 1.4 % relative standard deviation; duplicate plutonium determinations give a 0.3% relative standard deviation for sample analysis. There is a discrepancy (- 0.3% to - 0.9%) between the spectrophotometric method results and the isotopic dilution analysis [fr

  8. The role of troublesome components in plutonium vitrification

    Energy Technology Data Exchange (ETDEWEB)

    Li, Hong; Vienna, J.D.; Peeler, D.K.; Hrma, P.; Schweiger, M.J. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-05-01

    One option for immobilizing surplus plutonium is vitrification in a borosilicate glass. Two advantages of the glass form are (1) high tolerance to feed variability and, (2) high solubility of some impurity components. The types of plutonium-containing materials in the United States inventory include: pits, metals, oxides, residues, scrap, compounds, and fuel. Many of them also contain high concentrations of carbon, chloride, fluoride, phosphate, sulfate, and chromium oxide. To vitrify plutonium-containing scrap and residues, it is critical to understand the impact of each component on glass processing and chemical durability of the final product. This paper addresses glass processing issues associated with these troublesome components. It covers solubility limits of chlorine, fluorine, phosphate, sulfate, and chromium oxide in several borosilicate based glasses, and the effect of each component on vitrification (volatility, phase segregation, crystallization, and melt viscosity). Techniques (formulation, pretreatment, removal, and/or dilution) to mitigate the effect of these troublesome components are suggested.

  9. PREPARATION OF METAL OXIDE POWDERS FROM METAL LOADED VERSATIC ACID

    OpenAIRE

    KAKIHATA, Takayuki; USAMI, Kensuke; YAMAMOTO, Hideki; SHIBATA, Junji

    1998-01-01

    A production process for metal oxide powders was developed using a solvent extraction method. Versatic Acid 10 and D2EHPA solutions containing copper, zinc and nickel were used for a precipitation-stripping process, where oxalic acid was added to the solution as a precipitation reagent.Copper, zinc and nickel oxalates were easily formed in an aqueous phase, and 99.9% of precipitation was obtained for each metal during this process. These metal oxalates were easily converted to metal oxides by...

  10. 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.

  11. Experimental study of contamination by a mixture resulting from the combustion of sodium in the presence of plutonium oxide

    International Nuclear Information System (INIS)

    Metivier, H.; Masse, R.; Nenot, J.C.; Nolibe, D.; Lafuma, J.

    1976-01-01

    The introduction into industry of breeder reactors using liquid sodium as the coolant gives rise to fresh radiotoxicological problems; the most likely chemical form in which the plutonium and sodium mixture may be found after a possible accident is a soluble oxide of plutonium(VI) or plutonium(VII). The biological behaviour of such mixtures was analysed in rats and monkeys following parenteral administration. In the case of rats distribution in the organs was analysed from 30 minutes to 30 days following intramuscular contamination. In the monkeys the authors only analysed distribution in the organs after one month. The biological behaviour of the mixture is marked by very high plutonium solubility. The maximum burden in the organs attains roughly 20% of the quantity deposited both in monkeys and rats. The spread of the plutonium proceeds gradually from the day of administration; it indicates that a soluble form of plutonium is present for a long time. Also observed is an increasing osteotropic tendency in the diffusible form. The liver fraction is rapidly removed, if it is assumed that faecal excretion represents biliary excretion. In the experimental conditions selected, the spread of a liquid aerosol of the solution is rapid and constitutes about 10% of the quantity deposited in the airways at the end of inhalation. Treatment with DTPA makes it possible to reduce the body burden, but less efficiently than after contamination by Pu(IV). (author)

  12. Plutonium production and utilization forecasts in Europe

    International Nuclear Information System (INIS)

    Haijtink, B.

    1976-01-01

    The planned accelerated growth of nuclear energy generation in the near future will lead to a large production of plutonium in the thermal reactors. Therefore, up to 1985, the major part of the available plutonium will be plutonium recovered from spent uranium-metal, particularly in the United Kingdom and in France. Because of the low demand for fuelling the fast breeder reactors within the near future, a surplus of fissile plutonium will be accumulated in Europe. Even if the planned availability of the oxide reprocessing capacity will be delayed with two or three years, a plutonium surplus will still exist in Europe, e.g.; in 1985: 25-20 tons. On longer term, up to 2000, the plutonium production in thermal reactors will be sufficient to meet the estimated demand for fast breeder reactors at their commercial introduction foreseen for the nineties. That means that all the plutonium surplus needs not to be stocked for use in fast breeder reactors later on but could be recycled in thermal reactors. The magnitude of the available fissionable materials give an idea of the importance to promote, on an industrial scale, the plutonium recycling technology

  13. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

  14. Calculation of oxygen distribution in uranium-plutonium oxide fuels during irradiation (programme CODIF)

    International Nuclear Information System (INIS)

    Moreno, A.; Sari, C.

    1978-01-01

    Radial gradients of oxygen to metal ratio, O/M, in uranium-plutonium oxide fuel pins, during irradiation and at the end of life, have been calculated on the basis of solid-state thermal diffusion using measured values of the heat of transport. A detailed computer model which includes the calculation of temperature profiles and the variation of the average O/M ratio as a function of burn-up is given. Calculations show that oxygen profiles are affected by the isotopic composition of the fuel, by the temperature profiles and by fuel-cladding interactions

  15. An analytical method for the determination of plutonium in autopsy samples

    International Nuclear Information System (INIS)

    Santori, G.

    1983-01-01

    A sensitive method for the determination of plutonium in autopsy samples is described. After a suitable chemical pretreatment of the samples the plutonium is separated by extraction chromatography with tri-n-octylphosphine oxide (TOPO) supported on microporus polyethylene. After electrodeposition of plutonium the activity is counted by alpha spectroscopy. The global yield was 75-80%. The reagent blank activity was such to allow the determination of some femtocuries of plutonium

  16. Radiolytic gas generation in plutonium contaminated waste materials

    International Nuclear Information System (INIS)

    Kazanjian, A.R.

    1976-01-01

    Many plutonium contaminated waste materials decompose into gaseous products because of exposure to alpha radiation. The gases generated (usually hydrogen) over long-storage periods may create hazardous conditions. To determine the extent of such hazards, knowing the gas generation yields is necessary. These yields were measured by contacting some common Rocky Flats Plant waste materials with plutonium and monitoring the enclosed atmospheres for extensive periods of time. The materials were Plexiglas, polyvinyl chloride, glove-box gloves, machining oil, carbon tetrachloride, chlorothene VG solvent, Kimwipes (dry and wet), polyethylene, Dowex-1 resin, and surgeon's gloves. Both 239 Pu oxide and 238 Pu oxide were used as radiation sources. The gas analyses were made by mass spectrometry and the results obtained were the total gas generation, the hydrogen generation, the oxygen consumption rate, and the gas composition over the entire storage period. Hydrogen was the major gas produced in most of the materials. The total gas yields varied from 0.71 to 16 cm 3 (standard temperature pressure) per day per curie of plutonium. The oxygen consumption rates varied from 0.0088 to 0.070 millimoles per day per gram of plutonium oxide-239 and from 0.0014 to 0.0051 millimoles per day per milligram 238 Pu

  17. Fabrication of dendritic silver-coated copper powders by galvanic displacement reaction and their thermal stability against oxidation

    International Nuclear Information System (INIS)

    Park, Yu-Seon; An, Chang Yong; Kannan, Padmanathan Karthick; Seo, Nary; Zhuo, Kai; Yoo, Tae Kyong; Chung, Chan-Hwa

    2016-01-01

    Highlights: • The dendritic silver-coated copper powders with high specific surface area have been prepared using a simple wet chemical reduction process at room temperature. • It is found that the Cu starts to be oxidized into Cu_2O followed by CuO at elevated temperatures. • The more amount of Ag-coating provides the less oxidation, which confirms that the Ag-shell prevents the Cu-core from oxidation. • The resistivity of dendritic 33.27 wt.% Ag-coated Cu powders was measured to 25.67 μΩ cm after the annealing at 150 °C for 30 min. - Abstract: Two steps of wet chemical processes have been developed for the preparation of core-shell nanostructures of copper and silver, which is a facile and low cost method for the production of large quantity of dendritic powders. First step involves a galvanic displacement reaction with hydrogen evolution which is the motive force of spontaneous electrochemical reaction. To achieve the core-shell structure, silver has been coated on the dendritic copper using the galvanic displacement reaction. The dendritic silver-coated copper powders exhibit high surface-area, excellent conductivity, and good oxidation resistance. It has been found that silver-coated copper powders maintain the electrical conductivity even after annealing at 150 °C for several to tens of minutes, thus it is a promising material and an alternative to pure silver powders in printed electronics application.

  18. Fabrication of dendritic silver-coated copper powders by galvanic displacement reaction and their thermal stability against oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yu-Seon [School of Chemical Engineering, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Farad Materials Co., Ltd., Suwon 16419 (Korea, Republic of); An, Chang Yong; Kannan, Padmanathan Karthick; Seo, Nary [School of Chemical Engineering, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Zhuo, Kai [School of Chemical Engineering, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Farad Materials Co., Ltd., Suwon 16419 (Korea, Republic of); Yoo, Tae Kyong [School of Chemical Engineering, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Chung, Chan-Hwa, E-mail: chchung@skku.edu [School of Chemical Engineering, Sungkyunkwan University, Suwon 16419 (Korea, Republic of); Farad Materials Co., Ltd., Suwon 16419 (Korea, Republic of)

    2016-12-15

    Highlights: • The dendritic silver-coated copper powders with high specific surface area have been prepared using a simple wet chemical reduction process at room temperature. • It is found that the Cu starts to be oxidized into Cu{sub 2}O followed by CuO at elevated temperatures. • The more amount of Ag-coating provides the less oxidation, which confirms that the Ag-shell prevents the Cu-core from oxidation. • The resistivity of dendritic 33.27 wt.% Ag-coated Cu powders was measured to 25.67 μΩ cm after the annealing at 150 °C for 30 min. - Abstract: Two steps of wet chemical processes have been developed for the preparation of core-shell nanostructures of copper and silver, which is a facile and low cost method for the production of large quantity of dendritic powders. First step involves a galvanic displacement reaction with hydrogen evolution which is the motive force of spontaneous electrochemical reaction. To achieve the core-shell structure, silver has been coated on the dendritic copper using the galvanic displacement reaction. The dendritic silver-coated copper powders exhibit high surface-area, excellent conductivity, and good oxidation resistance. It has been found that silver-coated copper powders maintain the electrical conductivity even after annealing at 150 °C for several to tens of minutes, thus it is a promising material and an alternative to pure silver powders in printed electronics application.

  19. The role of the native oxide shell on the microwave sintering of copper metal powder compacts

    International Nuclear Information System (INIS)

    Mahmoud, Morsi M.; Link, Guido; Thumm, Manfred

    2015-01-01

    Highlights: • Thin oxide native layer had a critical role on microwave sintering of copper. • Explain why microwaves interact with copper powder differently than its bulk. • Abnormal expansion in copper is due to the plastic deformation and crack formation. • In-situ setup gives important insight about the microwave sintering of metals. • Microwave sintering is a promising candidate technology in powder metallurgy. - Abstract: Successful microwave sintering of several metal powders had been reported by many researchers with remarkable improvements in the materials properties and/or in the overall process. However, the concept behind microwave heating of metal powders has not been fully understood till now, as it is well known that bulk metals reflect microwaves. The progress of microwave sintering of copper metal powder compacts was investigated via combining both in-situ electrical resistivity and dilatometry measurements that give important information about microstructural changes with respect to the inter-particle electrical contacts during sintering. The sintering behavior of copper metal powders was depending on the type of the gas used, particle size, the initial green density, the soaking sintering time and the thin oxide layer on the particles surfaces. The thin copper oxide native layer (ceramics) that thermodynamically formed on the particles surfaces under normal handling and ambient environmental conditions had a very critical and important role in the microwave absorption and interaction, the sintering behavior and the microstructural changes. This finding could help to have a fundamental understanding of why MW’s interact with copper metal powder in a different way than its bulk at room temperature, i.e. why a given metal powder could be heated using microwaves while its bulk reflects it

  20. X-ray photoemission spectroscopy (XPS) study of uranium, neptunium and plutonium oxides in silicate-based glasses

    International Nuclear Information System (INIS)

    Lam, D.J.; Veal, B.W.; Paulikas, A.P.

    1982-11-01

    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

  1. Compositions of airborne plutonium-bearing particles from a plutonium finishing operation

    International Nuclear Information System (INIS)

    Sanders, S.M. Jr.

    1976-11-01

    The elemental composition of 111 plutonium-bearing particles was determined (using an electron microprobe) as part of a program to investigate the origin and behavior of the long-lived transuranic radionuclides released from fuel reprocessing facilities at the Savannah River Plant. These particles, collected from wet-cabinet and room-air exhausts from the plutonium finishing operation (JB-Line), were between 0.4 and 36 μm in diameter. Ninety-nine of the particles were found to be aggregates of various minerals and metals, six were quartz, and six were small (less than 2-μm-diameter) pieces of iron oxide. Collectively, these particles contained less minerals and more metals than natural dusts contain. The metallic constituents included elements normally not found in dusts, e.g., chromium, nickel, copper, and zinc. Concentrations of aluminum and iron exceeded those normally found in minerals. Elemental concentrations in individual particles covered a wide range: one 2-μm-diameter particle contained 97 percent NiO, a 9-μm-diameter particle contained 72 percent Cr 2 O 3 . Although the particles were selected because they produced plutonium fission tracks, the plutonium concentration was too low to be estimated by microprobe analysis in all but a 1-μm-diameter particle. This plutonium-bearing particle contained 73 percent PuO 2 by weight in combination with Fe 2 O 3 and mica; its activity was estimated at 0.17 pCi of 239 Pu

  2. Preparation of nano-iron oxide red pigment powders by use of cyanided tailings

    International Nuclear Information System (INIS)

    Li Dengxin; Gao Guolong; Meng Fanling; Ji Chong

    2008-01-01

    On one hand, cyanided tailings are one kind of pollutants. On the other hand, they contain a lot of valuable elements. So utilization of them can bring social and environmental benefits. In this paper, cyanided tailings were used to prepare nano-iron oxide red pigment powders by an ammonia process with urea as precipitant. At first, cyanided tailings were oxidized by nitric acid. Then, the oxidizing mixture was separated into solid and liquid parts. The liquid mixture was reduced by scrap iron and the impurity of it was removed by use of NH 3 .H 2 O. Then, the seed crystal of γ-FeOOH was obtained, when the pure liquid reacted with ammonia liquid at the selected experimental conditions. At last, nano-iron oxide red pigment powders were prepared. The structure, morphology and size distribution of seed crystal and iron oxide red were characterized systematically by means of X-ray diffraction (XRD), transmission electron microscope (TEM) and laser particle size analyzer (LPSA). The results revealed that typical iron oxide nanoparticles were α-Fe 2 O 3 with particle size of 50-70 nm. Furthermore, the factors that affected the hue and quality of the seed crystal and iron oxide red pigment were also discussed

  3. Radial plutonium redistribution in mixed-oxide fuel

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Schwinkendorf, K.N.; Karnesky, R.A.

    1981-10-01

    Alpha autoradiographs from all HEDL fuel pin metallography samples are evaluated and catalogued according to different plutonium distribution patterns. The data base is analyzed for effects of fabrication and operating parameters on redistribution

  4. Weapons-grade plutonium dispositioning. Volume 1: Executive summary

    International Nuclear Information System (INIS)

    Parks, D.L.; Sauerbrun, T.J.

    1993-06-01

    The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate dispositioning options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) assisted NAS in this evaluation by investigating the technical aspects of the dispositioning options and their capability for achieving plutonium annihilation levels greater than 90%. Additionally, the INEL investigated the feasibility of using plutonium fuels (without uranium) for disposal in existing light water reactors and provided a preconceptual analysis for a reactor specifically designed for destruction of weapons-grade plutonium. This four-volume report was prepared for NAS to document the findings of these studies. Volume 2 evaluates 12 plutonium dispositioning options. Volume 3 considers a concept for a low-temperature, low-pressure, low-power-density, low-coolant-flow-rate light water reactor that quickly destroys plutonium without using uranium or thorium. This reactor concept does not produce electricity and has no other mission than the destruction of plutonium. Volume 4 addresses neutronic performance, fabrication technology, and fuel performance and compatibility issues for zirconium-plutonium oxide fuels and aluminum-plutonium metallic fuels. This volumes gives summaries of Volumes 2--4

  5. Equilibrium-constant expressions for aqueous plutonium

    International Nuclear Information System (INIS)

    Silver, G.L.

    2010-01-01

    Equilibrium-constant expressions for Pu disproportionation reactions traditionally contain three or four terms representing the concentrations or fractions of the oxidation states. The expressions can be rewritten so that one of the oxidation states is replaced by a term containing the oxidation number of the plutonium. Experimental estimations of the numerical values of the constants can then be checked in several ways. (author)

  6. Plutonium Chemistry in the UREX Separation Processes

    International Nuclear Information System (INIS)

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

    2009-01-01

    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.

  7. The transport of plutonium industry in France. A high risk activity

    International Nuclear Information System (INIS)

    Coeytaux, X.; Faid, Y.B.; Hazemann, J.; Marignac, Y.; Schneider, M.

    2003-02-01

    According to the estimations of the report more than 450 transport of materials including about 40 tons of plutonium circulate in France during one year. The nuclear matters transported are a source of danger. The risks in relation with the handling and transport of plutonium touch the criticality risk, its great toxicity, and the problem of proliferation. To these risks it is necessary to add terrorists attacks. Three scenarios of accidents have been developed to illustrate the risk potential. A scenario of train accident, with derailment in a tunnel then crash with an other train; a simple scenario occurring to a lorry of plutonium powder; then a scenario of an accident following a terrorist act. These scenarios should be studied by I.R.S.N and the results should be given to the public and to elected people confronted to the transport of plutonium. (N.C.)

  8. Reaction of plutonium with water kinetic and equilibrium behavior of binary and ternary phases in the Pu + O + H system

    International Nuclear Information System (INIS)

    Haschke, J.M.; Hodges, A.E. III; Bixby, G.E.; Lucas, R.L.

    1983-01-01

    The kinetic and equilibrium behavior of the Pu + O + H system has been studied by measuring the production of hydrogen gas formed by a sequence of hydrolysis reactions. The kinetic dependence of the Pu + H 2 O reaction on salt concentration and temperature has been defined. The metal is quantitatively converted to a fine black powder which has been identified as plutonium monoxide monohydride, PuOH. Other hydrolysis products formed in aqueous media include a second oxide hydride, Pu 7 O 9 H 3 , and the oxides Pu 2 O 3 , Pu 7 O 12 , Pu 9 O 16 , Pu 10 O 18 , Pu 12 O 22 , and PuO 2 . Thermal decomposition products of PuOH include Pu 2 O 2 H and PuO. A tentative phase diagram for Pu + O + H is presented and structural relationships of the oxide hydrides and oxides are discussed. 10 figures, 5 tables

  9. Development of a pneumatic transport system for bulk transfer of metal grade uranium oxide powder

    International Nuclear Information System (INIS)

    Manna, S.; Satpati, S.K.; Roy, S.B.

    2010-01-01

    Uranium oxide powder is a commonly handled ceramic powder in nuclear industries. Design of the powder transfer system is an important aspect because of some of its typical characteristics. Pneumatic transport system has been widely used in transferring powder from one place to another. A pneumatic transport system using vacuum has been presented in the paper. This is used for bulk transfer of UO 3 powder. The system consists of a cyclone separator and filter cloth at the top of the cyclone separator. The pneumatic transfer system provides high efficiency with sustainable performance and it is a compact, robust, handy and moveable unit. No degradation of the powder quality has been observed during transfer. The system provides highly efficient, easy and safe transfer of radioactive powder, better working environment for the operator. (author)

  10. Review of major plutonium pyrochemical technology

    International Nuclear Information System (INIS)

    Moser, W.S.; Navratil, J.D.

    1983-01-01

    The past twenty years have seen significant growth in the development and application of pyrochemical technology for processing of plutonium. For particular feedstocks and specific applications, non-aqueous high-temperature processes offer key advantages over conventional hydrometallurgical systems. Major processes in use today include: (1) direct oxide reduction for conversion of PuO 2 to metal, (2) molten salt extraction for americium removal from plutonium, (3) molten salt electrorefining for Pu purification, and (4) hydriding to remove plutonium from host substrates. This paper reviews current major pyrochemical processes from the classical calcination-hydrofluorination-bomb reduction sequence through new techniques under development. Each process is presented and brief descriptions of production equipment are given. 47 references, 5 figures

  11. Synthesis of Uranium nitride powders using metal uranium powders

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Kim, Dong Joo; Oh, Jang Soo; Rhee, Young Woo; Kim, Jong Hun; Kim, Keon Sik

    2012-01-01

    Uranium nitride (UN) is a potential fuel material for advanced nuclear reactors because of their high fuel density, high thermal conductivity, high melting temperature, and considerable breeding capability in LWRs. Uranium nitride powders can be fabricated by a carbothermic reduction of the oxide powders, or the nitriding of metal uranium. The carbothermic reduction has an advantage in the production of fine powders. However it has many drawbacks such as an inevitable engagement of impurities, process burden, and difficulties in reusing of expensive N 15 gas. Manufacturing concerns issued in the carbothermic reduction process can be solved by changing the starting materials from oxide powder to metals. However, in nitriding process of metal, it is difficult to obtain fine nitride powders because metal uranium is usually fabricated in the form of bulk ingots. In this study, a simple reaction method was tested to fabricate uranium nitride powders directly from uranium metal powders. We fabricated uranium metal spherical powder and flake using a centrifugal atomization method. The nitride powders were obtained by thermal treating those metal particles under nitrogen containing gas. We investigated the phase and morphology evolutions of powders during the nitriding process. A phase analysis of nitride powders was also a part of the present work

  12. Long-term plutonium storage: Design concepts

    International Nuclear Information System (INIS)

    Wilkey, D.D.; Wood, W.T.; Guenther, C.D.

    1994-01-01

    An important part of the Department of Energy (DOE) Weapons Complex Reconfiguration (WCR) Program is the development of facilities for long-term storage of plutonium. The WCR design goals are to provide storage for metals, oxides, pits, and fuel-grade plutonium, including material being held as part of the Strategic Reserve and excess material. Major activities associated with plutonium storage are sorting the plutonium inventory, material handling and storage support, shipping and receiving, and surveillance of material in storage for both safety evaluations and safeguards and security. A variety of methods for plutonium storage have been used, both within the DOE weapons complex and by external organizations. This paper discusses the advantages and disadvantages of proposed storage concepts based upon functional criteria. The concepts discussed include floor wells, vertical and horizontal sleeves, warehouse storage on vertical racks, and modular storage units. Issues/factors considered in determining a preferred design include operational efficiency, maintenance and repair, environmental impact, radiation and criticality safety, safeguards and security, heat removal, waste minimization, international inspection requirements, and construction and operational costs

  13. Comparison of NDA and DA measurement techniques for excess Pu powders at the Hanford Site: Operator and IAEA experience

    International Nuclear Information System (INIS)

    Welsh, T.L.; McRae, L.P.; Delegard, C.H.

    1995-06-01

    Quantitative physical measurements are necessary components of the International Atomic Energy Agency (IAEA) nuclear material safeguards verification regime. In December 1994, IAEA safeguards were initiated on an inventory of plutonium-bearing oxide and scrap items in Vault 3 of the 2736-Z Building of the Plutonium Finishing Plant on the United States Department of Energy's (USDOE) Hanford Site. The material originated in the United States nuclear weapons complex. The diversity of the chemical form and the heterogenous physical form of the plutonium in this inventory were expected to challenge the target precision and accuracy of methods employed by IAEA: quantitative destructive analytical techniques (which are susceptible to sampling error) and quantitative coincident neutron measurements (which rely on knowledge of the material's chemical form and purity). Because of the diverse and heterogenous nature of plutonium-bearing scrap, plant operations increasingly have adopted calorimetric techniques both for item inventory measurements and for verification purposes. During the recent advent of IAEA safeguards at Vault 3, a set of destructive and nondestructive methods were applied to a number of inventory items (cans of plutonium-bearing powders) with widely ranging chemical purities. Results of these measurements, gathered by the operator's and IAEA's laboratories and instruments as well as by instruments from Pacific Northwest Laboratory and USDOE's Los Alamos National Laboratory (LANL), are presented and statistically compared

  14. Plutonium stabilization and handling quality assurance program plan

    International Nuclear Information System (INIS)

    Weiss, E.V.

    1998-01-01

    This Quality Assurance Program Plan (QAPP) identifies project quality assurance requirements for all contractors involved in the planning and execution of Hanford Site activities for design, procurement, construction, testing and inspection for Project W-460, Plutonium Stabilization and Handling. The project encompasses procurement and installation of a Stabilization and Packaging System (SPS) to oxidize and package for long term storage remaining plutonium-bearing special nuclear materials currently in inventory at the Plutonium Finishing Plant (PFP), and modification of vault equipment to allow storage of resulting packages of stabilized SNM

  15. Risk of transporting plutonium dioxide and liquid plutonium nitrate by truck and rail

    International Nuclear Information System (INIS)

    Williams, L.D.; Hall, R.J.

    1976-01-01

    This paper presents an analysis method developed to assess the risk in the transportation of radioactive materials and demonstrates its application by summarizing the results of risk assessments of the shipment of plutonium dioxide powder and liquid plutonium nitrate by truck and rail in the United States. In the risk assessment, postulated release modes (series of events that could result in a release of radioactive material) are identified using the fault tree analysis method. Two categories of events that could contribute to a release during transportation are considered: accidents and substandard packaging conditions. The likelihood of basic events in these categories are determined from general accident experience, radioactive material shipping experience, engineering analysis, etc. Using the laws of combining probabilities, the likelihood of radioactive material release is then evaluated. Since accidents vary greatly in severity, the releases from these postulated events also vary in magnitude from the no release condition up through release of the entire package contents. The bases for the release quantities assigned to the postulated accidents sequences are discussed. The consequences of these postulated releases are evaluated based on the amount of plutonium estimated to be released to the environs, the probable weather conditions at the time of the accident, and population density downwind from the accident scene. The likelihood and consequence of postulated releases are coupled and expressed as a risk spectra. (author)

  16. Speciation of plutonium during sorption and diffusion in Opalinus clay

    International Nuclear Information System (INIS)

    Kaplan, Ugras

    2013-01-01

    The presented work was carried out in the framework of the BMWi-project ''Interaction and migration of actinides in natural clay rocks taking into account humic substances and clay organic matter - Interactions of neptunium and plutonium with natural clay rocks''. For the long-term safety assessments of nuclear repositories, the possible migration of the radiotoxic wastes into the environment must be considered. Due to its long half-life (T 1/2 = 24000 y) 239 Pu has a significant contribution to the radiotoxicity of spent nuclear fuel in a repository after long periods of storage. The redox-sensitive plutonium has a very complicated chemical behavior. In aqueous solution under environmental relevant conditions Pu can exist in oxidation states +III to +VI and it can exist in up to four oxidation states simultaneously in a solution. Clays are considered as a possible host rock formation for of high-level radioactive waste disposal. Therefore, detailed information on the mobilization and immobilization of plutonium through / into the groundwater from a repository are of special interest. In this work new insights into the interaction between Pu and natural Opalinus clay (OPA, Mont Terri, Switzerland) are obtained with regard to the disposal of heat-generating radioactive waste in a deep geological repository.rnThe focus of this work was on the determination of the speciation of Pu on the mineral surface after sorption and diffusion process by different synchrotron based techniques (μ-XRF, μ-XANES/-EXAFS, μ-XRD, and EXAFS/XANES). The interaction between Pu and OPA was studied in batch sorption and diffusion experiments in dependence of various experimental parameters (e.g. pH, Pu oxidation state). Sorption experiments showed that some experimental parameters (e.g. temperature, humic acid) have a significant impact on the sorption of Pu on OPA. Speciation studies were performed as a function of various chemical parameters on powder samples form batch experiments as

  17. 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.

  18. Pyrochemical recovery of plutonium from calcium fluoride reduction slag

    Science.gov (United States)

    Christensen, D.C.

    A pyrochemical method of recovering finely dispersed plutonium metal from calcium fluoride reduction slag is claimed. The plutonium-bearing slag is crushed and melted in the presence of at least an equimolar amount of calcium chloride and a few percent metallic calcium. The calcium chloride reduces the melting point and thereby decreases the viscosity of the molten mixture. The calcium reduces any oxidized plutonium in the mixture and also causes the dispersed plutonium metal to coalesce and settle out as a separate metallic phase at the bottom of the reaction vessel. Upon cooling the mixture to room temperature, the solid plutonium can be cleanly separated from the overlying solid slag, with an average recovery yield on the order of 96 percent.

  19. Preparation standardisation and use of plutonium nitrate reference solutions

    International Nuclear Information System (INIS)

    Brown, M.L.; Drummond, J.L.

    1981-07-01

    A procedure is described for the purification of a plutonium nitrate solution in nitric acid for use as a plutonium master standard. Anion exchange chromatography followed by oxalate precipitation is used to purify the plutonium and the residual cationic impurities are analysed by emission spectroscopy. The plutonium content is accurately and precisely measured by two independent methods, namely by gravimetry as PuO 2 at 1250 0 C and by ceric oxidation, ferrous reduction and dichromate titration. Full details of the purification procedure are given, with recommended methods for storing and using the standard solution. It is concluded that such a solution is the most satisfactory reference material, available for plutonium analysis for reprocessing plants, and is adequately related to other, internationally accepted, standard reference materials. (author)

  20. Study of the dissolution of (U,P)O2 mixed oxides with a high plutonium content

    International Nuclear Information System (INIS)

    Fournier, S.

    2001-01-01

    Plutonium from nuclear reactors is partially integrated in the fuel cycle as Mixed OXide (U,Pu)O 2 (MOX). Their dissolution in nitric acid is needed to reprocess them in present reprocessing plants. The main difficulty of this study is that dissolution is a phenomenon depending on solution characteristics as well as the structural properties of the pellets, which depend themselves on the material fabrication process. After showing kinetic and thermodynamic dissolution data of mixed oxides in nitric media, an inventory of the parameters which affect the dissolution process has been made. A separable variable concept was introduced in order to describe the process by studying separately the role of chemical parameters of the solution and geometric parameters of the material. The first part of the study estimates the effect of nitric solution chemical parameters (concentrations, acidity, temperature) on the dissolution and underlines the role of the oxide surface protonation step. The second part of this work deals with the study of surface area evolution for materials with controlled plutonium rich heterogeneities. Experimental results show that the pellet surface undergoes erosion and is progressively weakened by the formation of fault lines in the bulk of the material followed by the dispersion of sub-millimeter fragments in the solution. An heterogeneous kinetic model derived from study of solid-gas interface systems has been applied to fuel pellets dissolution, allowing a mechanism to be proposed, based on surface dissolution of the oxide as well as fault creation in the volume. The dissolution kinetics are therefore dependant on the microstructure and mechanical strength and cohesion of the pellets. (author)

  1. Overview of chemical characterization of FBTR fuel

    International Nuclear Information System (INIS)

    Venkatesan, V.; Nandi, C.; Patil, A.B.; Prakash, Amrit; Khan, K.B.; Arun Kumar

    2015-01-01

    Uranium Plutonium mixed carbide fuel is the driver fuel for Fast Breeder Test Reactor (FBTR) at IGCAR. The fuel is being fabricated at Radiometallurgy Division, BARC by conventional powder metallurgy route. During the fabrication of fuel, chemical quality control of process intermediates is very important to reach stringent specification of the final fuel product. Different steps are involved in the fabrication of uranium-plutonium carbide (MC) for FBTR. The main steps in the fabrication of MC fuel pellets are carbothermic reduction (CR) of mixture of uranium oxide, plutonium oxide and graphite powder to prepare MC clinkers, crushing and milling of MC clinkers and consolidation of MC powders into fuel pellets and sintering. As a part of process control, analysis of uranium (U), plutonium (Pu), carbon in oxide graphite mixture and U, Pu, carbon, oxygen, nitrogen, MC, M 2 C 3 contents in mixed carbide powder (MC clinkers) are carried out at our laboratory. Analysis of U, Pu, carbon, oxygen, nitrogen, MC and M 2 C 3 contents in mixed carbide sintered pellets are carried out as a part of quality control. This paper describes an overview of analytical instruments used during chemical quality control of mixed carbide fuel

  2. Optimal management of weapons plutonium through MOX recycling

    International Nuclear Information System (INIS)

    McMurphy, M.A.; Bastard, G. le

    1995-01-01

    Beyond the satisfaction of witnessing the end of the nuclear arms race, the availability of large quantities of plutonium from the dismantlement of nuclear weapons in Russia and the US can be perceived as a challenge and an opportunity. A challenge because poor management of this material would maintain a problematic situation in terms of proliferation; an opportunity because such plutonium represents a high value energy source that the civilian industry is capable of using efficiently, actually turning it from swords to plowshares. The object of this paper is to describe the main characteristics of the use of weapons plutonium in the civilian cycle to produce electricity through the use of mixed uranium-plutonium oxide (MOX), or moxification. A comparison with the main alternate solution--plutonium vitrification--is offered, in particular with regard to industrial availability, energy resource management, economy, environment and proliferation

  3. Depositing laser-generated nanoparticles on powders for additive manufacturing of oxide dispersed strengthened alloy parts via laser metal deposition

    Science.gov (United States)

    Streubel, René; Wilms, Markus B.; Doñate-Buendía, Carlos; Weisheit, Andreas; Barcikowski, Stephan; Henrich Schleifenbaum, Johannes; Gökce, Bilal

    2018-04-01

    We present a novel route for the adsorption of pulsed laser-dispersed nanoparticles onto metal powders in aqueous solution without using any binders or surfactants. By electrostatic interaction, we deposit Y2O3 nanoparticles onto iron-chromium based powders and obtain a high dispersion of nano-sized particles on the metallic powders. Within the additively manufactured component, we show that the particle spacing of the oxide inclusion can be adjusted by the initial mass fraction of the adsorbed Y2O3 particles on the micropowder. Thus, our procedure constitutes a robust route for additive manufacturing of oxide dispersion-strengthened alloys via oxide nanoparticles supported on steel micropowders.

  4. Physics of Plutonium Recycling in Thermal Reactors

    International Nuclear Information System (INIS)

    Kinchin, G.H.

    1967-01-01

    A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of 240 Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

  5. Physics of Plutonium Recycling in Thermal Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kinchin, G. H. [Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1967-09-15

    A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of {sup 240}Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

  6. Adsorption of neptunium and plutonium on metal phosphites

    International Nuclear Information System (INIS)

    Silver, G.L.

    1979-01-01

    The removal of neptunium and plutonium from water by adsorption on titanium, zirconium, bismuth, thorium, and uranium phosphites was investigated. These phosphites hydrolyze in neutral or alkaline solution producing the hydrous metal oxides that are more effective adsorbents than the original phosphite compounds. Ageing the plutonium-238 polymer changes its adsorption characteristics on commercial bone char. 37 figures, 7 tables

  7. Swiss R and D on uranium-free LWR fuels for plutonium incineration

    International Nuclear Information System (INIS)

    Stanculescu, A.; Chawla, R.; Degueldre, C.; Kasemeyer, U.; Ledergerber, G.; Paratte, J.M.

    1999-01-01

    The most efficient way to enhance the plutonium consumption in LWRs is to eliminate plutonium production altogether. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. The inert matrix material studied at PSI is zirconium oxide. For reactivity control reasons, adding a burnable poison to this fuel proves to be necessary. The studies performed at PSI have identified erbium oxide as the most suitable candidate for this purpose. With regard to material technology aspects, efforts have concentrated on the evaluation of fabrication feasibility and on the determination of the physicochemical properties of the chosen single phase zirconium/ erbium/plutonium oxide material stabilised as a cubic solution by yttrium. The results to-date, obtained for inert matrix samples containing thorium or cerium as plutonium substitute, confirm the robustness and stability of this material. With regard to reactor physics aspects, our studies indicate the feasibility of uranium-free, plutonium-fuelled cores having operational characteristics quite similar to those of conventional UO 2 -fuelled ones, and much higher plutonium consumption rates, as compared to 100% MOX loadings. The safety features of such cores, based on results obtained from static neutronics calculations, show no cliff edges. However, the need for further detailed transient analyses is clearly recognised. Summarising, PSI's studies indicate the feasibility of a uranium-free plutonium fuel to be considered in 'maximum plutonium consumption LWRs' operating in a 'once-through' mode. With regard to reactor physics, future efforts will concentrate on strengthening the safety case of uranium-free cores, as well as on improving the integral data base for validation of the neutronics calculations. Material technology studies will be continued to investigate the physico-chemical properties of the inert matrix fuel containing plutonium and will focus on the planning and evaluation of

  8. X-ray absorption near-edge spectroscopy of plutonium solid species

    International Nuclear Information System (INIS)

    Kropf, A. J.

    1998-01-01

    We present XANES at the L III edge for four plutonium solid phases: Pu(III)F 3 , Pu(IV)O 2 , NaPu(V)O 2 CO 3 , and Ba 3 Pu(VI)O 6 . These correspond to the four important oxidation states in the process chemistry and environmental chemistry of plutonium. By a fitting method that uses an arc tangent function and gaussian curves, it was possible to reproducibly determine the edge energy and distinguish among the four oxidation states. These data demonstrate a 1.85 ± 0.20 eV shift per oxidation state

  9. Redox reaction between graphene oxide and In powder to prepare In2O3/reduced graphene oxide hybrids for supercapacitors

    Science.gov (United States)

    Xu, Xiaoyang; Wu, Tao; Xia, Fengling; Li, Yi; Zhang, Congcong; Zhang, Lei; Chen, Mingxi; Li, Xichuan; Zhang, Li; Liu, Yu; Gao, Jianping

    2014-11-01

    A facile and quick route for the chemical reduction of graphene oxide (GO) using In powder as a reductant has been established. The reduction of GO by In powder is traced by UV-visible absorption spectroscopy, and the obtained reduced graphene oxide (rGO) is analyzed. The In3+ ions produced during the reaction between the GO and the In powder are chemically transformed to In2O3 and then form In2O3/rGO hybrids. The In2O3/rGO hybrids are used as electrode materials and their electrochemical performance are studied using cyclic voltammetry and galvanostatic charge/discharge. The In2O3/rGO hybrids demonstrate excellent electrochemical performance and their highest specific capacitance is 178.8 F g-1 which is much higher than that of either In2O3 or rGO. In addition, the In2O3/rGO hybrids are also very stable.

  10. Technique of preparation of plutonium screens for soft x ray spectrography (1963

    International Nuclear Information System (INIS)

    Bersuder, L. de

    1963-01-01

    The present work concerns the preparation of thin layers of pure plutonium (thickness 100 to 10000 A) by thermal vacuum evaporation. The protection of the plutonium against oxidation is obtained by vacuum deposition of aluminium layers under. and above the plutonium layer. The purity of the layers is checked by electron and X ray diffraction which has shown that very thin films of plutonium condense in β form instead of α. (author) [fr

  11. Robotic sample preparation for radiochemical plutonium and americium analyses

    International Nuclear Information System (INIS)

    Stalnaker, N.; Beugelsdijk, T.; Thurston, A.; Quintana, J.

    1985-01-01

    A Zymate robotic system has been assembled and programmed to prepare samples for plutonium and americium analyses by radioactivity counting. The system performs two procedures: a simple dilution procedure and a TTA (xylene) extraction of plutonium. To perform the procedures, the robotic system executes 11 unit operations such as weighing, pipetting, mixing, etc. Approximately 150 programs, which require 64 kilobytes of memory, control the system. The system is now being tested with high-purity plutonium metal and plutonium oxide samples. Our studies indicate that the system can give results that agree within 5% at the 95% confidence level with determinations performed manually. 1 ref., 1 fig., 1 tab

  12. Physics of plutonium recycling

    International Nuclear Information System (INIS)

    2003-01-01

    The commercial recycling of plutonium as PuO 2 /UO 2 mixed-oxide (MOX) fuel is an established practice in pressurised water reactors (PWRs) in several countries, the main motivation being the consumption of plutonium arising from spent fuel reprocessing. Although the same motivating factors apply in the case of boiling water reactors (BWRs), they have lagged behind PWRs for various reasons, and MOX utilisation in BWRs has been implemented in only a few reactors to date. One of the reasons is that the nuclear design of BWR MOX assemblies (or bundles) is more complex than that of PWR assemblies. Recognizing the need and the timeliness to address this issue at the international level, the OECD/NEA Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles (WPPR) conducted a physics code benchmark test for a BWR assembly. This volume reports on the benchmark results and conclusions that can be drawn from it. (authors)

  13. Disposal of Surplus Weapons Grade Plutonium

    International Nuclear Information System (INIS)

    Alsaed, H.; Gottlieb, P.

    2000-01-01

    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

  14. Development of the conductive glove for glove box operation. Production of prototypes and their experimental results

    International Nuclear Information System (INIS)

    Kodato, Kazuo; Enuma, Masahito; Kawasaki, Takeshi; Nogami, Yoshitaka; Kaneko, Kazunori; Kimura, Masanori; Yasumori, Tomokazu

    2014-02-01

    The glove used at glove boxes in the nuclear fuel plants is usually made with Chlorosulfonated polyethylene rubber. The rubber is excellent in terms of resistance to radiation because it has no double bond in its main chain of the component, however, it deteriorates rapidly in high dose environment such as direct contact of alpha ray. Plutonium oxide powder is treated in glove boxes at plutonium fuel facilities where the alpha ray from plutonium oxide powder adhered on surface of a glove causes the deterioration of the rubber. Therefore the effective method for prevent of the rapid deterioration is to decrease the amount of adhered powder, and the glove with conductive property which can prevent static electric charge on its surface has been developed and tested. The results showed that the rubber has less adherent property to powder compared with conventional one. (author)

  15. PROCESS USING POTASSIUM LANTHANUM SULFATE FOR FORMING A CARRIER PRECIPITATE FOR PLUTONIUM VALUES

    Science.gov (United States)

    Angerman, A.A.

    1958-10-21

    A process is presented for recovering plutonium values in an oxidation state not greater than +4 from fluoride-soluble fission products. The process consists of adding to an aqueous acidic solution of such plutonium values a crystalline potassium lanthanum sulfate precipitate which carries the plutonium values from the solution.

  16. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    International Nuclear Information System (INIS)

    Wishau, R.; Ramsey, K.B.; Montoya, A.

    1998-01-01

    This paper presents the technical and economic feasibility of molten salt oxidation technology as a volume reduction and recovery process for 238 Pu contaminated waste. Combustible low-level waste material contaminated with 238 Pu residue is destroyed by oxidation in a 900 C molten salt reaction vessel. The combustible waste is destroyed creating carbon dioxide and steam and a small amount of ash and insoluble 2328 Pu in the spent salt. The valuable 238 Pu is recycled using aqueous recovery techniques. Experimental test results for this technology indicate a plutonium recovery efficiency of 99%. Molten salt oxidation stabilizes the waste converting it to a non-combustible waste. Thus installation and use of molten salt oxidation technology will substantially reduce the volume of 238 Pu contaminated waste. Cost-effectiveness evaluations of molten salt oxidation indicate a significant cost savings when compared to the present plans to package, or re-package, certify and transport these wastes to the Waste Isolation Pilot Plant for permanent disposal. Clear and distinct cost advantages exist for MSO when the monetary value of the recovered 238 Pu is considered

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

    International Nuclear Information System (INIS)

    Subba Rao, R.V.; Damodaran, K.; Santosh Kumar, G.; Ravi, T.N.

    2000-01-01

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

  18. Reduction of worldwide plutonium inventories using conventional reactors and advanced fuels: A systems study

    International Nuclear Information System (INIS)

    Krakowski, R.A.; Bathke, C.G.; Chodak, P. III

    1997-01-01

    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 that includes an approximate measure of proliferation risk. The impact of plutonium recycle in a number of forms is examined, including the introduction of nonfertile fuels into conventional (LWR) reactors to reduce net plutonium generation, to increase plutonium burnup, and to reduce exo-reactor plutonium inventories

  19. 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.

  20. Specification for nuclear-grade beryllium oxide powder

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification defines the physical and chemical requirements of nuclear-grade beryllium oxide (BeO) powder to be used in fabricating nuclear components. 1.2 This specification does not include requirements for health and safety. , , It recognizes the material as a Class B poison and suggests that producers and users become thoroughly familiar with and comply to applicable federal, state, and local regulations and handling guidelines. 1.3 Special tests and procedures are given in Annex A1 and Annex A2. 1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  1. Characterization of manganese-gallium mixed oxide powders

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez Escribano, V.; Fernandez Lopez, E.; Sanchez Huidobro, P. [Universidad de Salamanca, Dept. de Quimica Inorganica (Spain); Panizza, M.; Resini, C.; Busca, G. [UNiversita di Genova, Dipt. di Ingegneria Chimica e di Processo, Genova (Italy); Resini, C. [Istituto Nazionale di Fisica della Materia, INFM (Spain); Gallardo- Amores, J.M. [Universidad Complutense, Dept. de Quimica Inorganica, Lab. Complutense de Altas Presiones, Madrid (Spain)

    2003-12-01

    Mn-Ga mixed oxides have been prepared by coprecipitation of the corresponding oxo-hydroxides as powders and have been characterized in relation to their structural and optical properties. The materials have been characterized by XRD, TG-DTA, skeletal IR and UV-visible-NIR spectroscopies. Large solubility of Mn in the diaspore type {alpha}-GaOOH oxo-hydroxide has been found. The spinel related structures of hausmannite Mn{sub 3}O{sub 4} and of {beta}-gallia present large reciprocal solubilities at least in a metastable form. At high temperature also bixbyite-type {alpha}-Mn{sub O3} solid solutions containing up to 20% at. Ga have been observed. (authors)

  2. Characterization of manganese?gallium mixed oxide powders

    Science.gov (United States)

    Sánchez Escribano, Vicente; Fernández López, Enrique; Sánchez Huidobro, Paula; Panizza, Marta; Resini, Carlo; Gallardo-Amores, José M.; Busca, Guido

    2003-11-01

    MnGa mixed oxides have been prepared by coprecipitation of the corresponding oxo-hydroxides as powders and have been characterized in relation to their structural and optical properties. The materials have been characterized by XRD, TG-DTA, skeletal IR and UV-visible-NIR spectroscopies. Large solubility of Mn in the diaspore type α-GaOOH oxo-hydroxide has been found. The spinel related structures of hausmannite Mn 3O 4 and of β-gallia present large reciprocal solubilities at least in a metastable form. At high temperature also bixbyite-type α-Mn 2O 3 solid solutions containing up to 20% at. Ga have been observed.

  3. Surplus plutonium disposition draft environmental impact statement. Summary

    International Nuclear Information System (INIS)

    1998-07-01

    On May 22, 1997, DOE published a Notice of Intent (NOI) in the Federal Register (62 Federal Register 28009) announcing its decision to prepare an environmental impact statement (EIS) that would tier from the analysis and decisions reached in connection with the Storage and Disposition of Weapons-Usable Fissile Materials Final Programmatic EIS (Storage and Disposition PEIS). DOE's disposition strategy allows for both the immobilization of surplus plutonium and its use as mixed oxide (MOX) fuel in existing domestic, commercial reactors. The disposition of surplus plutonium would also involve disposal of the immobilized plutonium and MOX fuel (as spent nuclear fuel) in a geologic repository. The Surplus Plutonium Disposition Environmental Impact Statement analyzes alternatives that would use the immobilization approach (for some of the surplus plutonium) and the MOX fuel approach (for some of the surplus plutonium); alternatives that would immobilize all of the surplus plutonium; and the No Action Alternative. The alternatives include three disposition facilities that would be designed so that they could collectively accomplish disposition of up to 50 metric tons (55 tons) of surplus plutonium over their operating lives: (1) the pit disassembly and conversion facility would disassemble pits (a weapons component) and convert the recovered plutonium, as well as plutonium metal from other sources, into plutonium dioxide suitable for disposition; (2) the immobilization facility would include a collocated capability for converting nonpit plutonium materials into plutonium dioxide suitable for immobilization and would be located at either Hanford or SRS. DOE has identified SRS as the preferred site for an immobilization facility; (3) the MOX fuel fabrication facility would fabricate plutonium dioxide into MOX fuel

  4. Prototype fast neutron counter for the assay of impure plutonium

    International Nuclear Information System (INIS)

    Wachter, J.R.; Adams, E.L.; Ensslin, N.

    1987-01-01

    A fast coincident neutron counter using liquid scintillators and gamma-ray/neutron pulse-shape discrimination has been constructed for the analysis of plutonium samples with unknown self-multiplication and (α,n) production. The counter was used to measure plutonium-bearing materials that cover a range of masses and (α,n) reaction rates of importance to the safeguards community. Measured values of the 240 Pu effective mass differed, on average, from their declared values by 0.4% for plutonium oxides and by -2.2% for metal and MgO-loaded samples. Poorer results were obtained for materials with large (α,n) reaction rates and low self-multiplication such as plutonium ash and plutonium fluoride

  5. Manufacturing And High Temperature Oxidation Properties Of Electro-Sprayed Fe-24.5% Cr-5%Al Powder Porous Metal

    Directory of Open Access Journals (Sweden)

    Lee Kee-Ahn

    2015-06-01

    Full Text Available Fe-Cr-Al based Powder porous metals were manufactured using a new electro-spray process, and the microstructures and high-temperature oxidation properties were examined. The porous materials were obtained at different sintering temperatures (1350°C, 1400°C, 1450°C, and 1500°C and with different pore sizes (500 μm, 450 μm, and 200 μm. High-temperature oxidation experiments (TGA, Thermal Gravimetry Analysis were conducted for 24 hours at 1000°C in a 79% N2+ 21% O2, 100 mL/min. atmosphere. The Fe-Cr-Al powder porous metals manufactured through the electro-spray process showed more-excellent oxidation resistance as sintering temperature and pore size increased. In addition, the fact that the densities and surface areas of the abovementioned powder porous metals had the largest effects on the metal’s oxidation properties could be identified.

  6. Installation and performance evaluation of an indigenous surface area analyser

    International Nuclear Information System (INIS)

    Pillai, S.N.; Solapurkar, M.N.; Venkatesan, V.; Prakash, A.; Khan, K.B.; Kumar, Arun; Prasad, R.S.

    2014-01-01

    An indigenously available surface area analyser was installed inside glove box and checked for its performance by analyzing uranium oxide and thorium oxide powders at RMD. The unit has been made ready for analysis of Plutonium oxide powders after incorporating several important features. (author)

  7. Anti-oxidant effect of extracts of kinnow rind, pomegranate rind and seed powders in cooked goat meat patties.

    Science.gov (United States)

    Devatkal, Suresh K; Narsaiah, K; Borah, A

    2010-05-01

    To overcome the disadvantages of using synthetic anti-oxidants in meat products, an investigation was carried out to evaluate the anti-oxidant effect of extracts of fruit by-products viz., kinnow rind powder (KRP), pomegranate rind powder (PRP) and pomegranate seed powder (PSP) in goat meat patties. Total phenolics content, DPPH radical scavenging activity and effect of these extracts on instrumental color, sensory attributes and TBARS values during storage (4+/-1 degrees C) of goat meat patties were evaluated. Results showed that these extracts are rich sources of phenolic compounds having free radical scavenging activity. Hunter Lab L value significantly (Pgoat meat patties was observed in PRP, PSP and KRP as compared to control patties. Average TBARS values (mg/kg meat) during refrigerated storage (4+/-1 degrees C) were significantly lower in PRP, followed by PSP and KRP as compared to control. The overall anti-oxidant effect was in the order of PRP>PSP>KRP. It was concluded that extracts of above fruits by-product powders have potential to be used as natural anti-oxidants in meat products. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

  8. Conversion of plutonium scrap and residue to boroilicate glass using the GMODS process

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.; Rudolph, J.; Elam, K.R.; Ferrada, J.J.

    1995-01-01

    Plutonium scrap and residue represent major national and international concerns because (1) significant environmental, safety, and health (ES ampersand H) problems have been identified with their storage; (2) all plutonium recovered from the black market in Europe has been from this category; (3) storage costs are high; and (4) safeguards are difficult. It is proposed to address these problems by conversion of plutonium scrap and residue to a CRACHIP (CRiticality, Aerosol, and CHemically Inert Plutonium) glass using the Glass Material Oxidation and Dissolution System (GMODS). CRACHIP refers to a set of requirements for plutonium storage forms that minimize ES ampersand H concerns. The concept is several decades old. Conversion of plutonium from complex chemical mixtures and variable geometries into a certified, qualified, homogeneous CRACHIP glass creates a stable chemical form that minimizes ES ampersand H risks, simplifies safeguards and security, provides an easy-to-store form, decreases storage costs, and allows for future disposition options. GMODS is a new process to directly convert metals, ceramics, and amorphous solids to glass; oxidize organics with the residue converted to glass; and convert chlorides to borosilicate glass and a secondary sodium chloride stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium (a plutonium surrogate), Zircaloy, stainless steel, and other materials to glass. GMODS is an enabling technology that creates new options. Conventional glassmaking processes require conversion of feeds to oxide-like forms before final conversion to glass. Such chemical conversion and separation processes are often complex and expensive

  9. EIS Data Call Report: Plutonium immobilization plant using ceramic in new facilities at the Savannah River Site

    International Nuclear Information System (INIS)

    DiSabatino, A.

    1998-01-01

    The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. This immobilization process is shown conceptually in Figure 1-1. The objective is to make an immobilized form, suitable for geologic disposal, in which the plutonium is as inherently unattractive and inaccessible as the plutonium in spent fuel from commercial reactors. The ceramic immobilization alternative presented in this report consists of first converting the surplus material to an oxide, followed by incorporating the plutonium oxide into a titanate-based ceramic material that is placed in metal cans

  10. Characterization of uranium- and plutonium-contaminated soils by electron microscopy

    International Nuclear Information System (INIS)

    Buck, E.C.; Dietz, N.L.; Fortner, J.A.; Bates, J.K.; Brown, N.R.

    1995-01-01

    Electron beam techniques have been used to characterize uranium-contaminated soils from the Fernald Site in Ohio, and also plutonium-bearing 'hot particles, from Johnston Island in the Pacific Ocean. By examining Fernald samples that had undergone chemical leaching it was possible to observe the effect the treatment had on specific uranium-bearing phases. The technique of Heap leaching, using carbonate solution, was found to be the most successful in removing uranium from Fernald soils, the Heap process allows aeration, which facilitates the oxidation of uraninite. However, another refractory uranium(IV) phase, uranium metaphosphate, was not removed or affected by any soil-washing process. Examination of ''hot particles'' from Johnston Island revealed that plutonium and uranium were present in 50--200 nm particles, both amorphous and crystalline, within a partially amorphous aluminum oxide matrix. The aluminum oxide is believed to have undergone a crystalline-to-amorphous transition caused by alpha-particle bombardment during the decay of the plutonium

  11. US and Russia face urgent decisions on weapons plutonium

    International Nuclear Information System (INIS)

    Hileman, B.

    1994-01-01

    Surplus plutonium poses a ''clear and present danger to national and international security,'' warns a National Academy of Sciences (NAS) study released in January, titled ''The Management and Disposition of Excess Weapons Plutonium.'' Over the past few years, many different methods of disposing of plutonium have been proposed. They range from shooting it into the Sun with missiles, to deep-seabed disposal, to fissioning it within a new generation of nuclear reactors. The NAS report rejects most of the methods suggested so far, but does recommend pursuing two of the options. One is to incorporate the plutonium in mixed-oxide fuel, a mixture of plutonium and uranium oxides, and use it to fuel commercial nuclear reactors. The other is to mix the plutonium with high-level waste and molten glass and mold the resulting material into large glass logs for eventual geologic disposal. Both are discussed here. The panel that wrote the NAS study is a standing committee called the Committee on International Security ampersand Arms Control. It suggests steps that should be taken now to guard supplies of plutonium removed from weapons. One step is bilateral US-Russian monitoring of warhead dismantlement. Others include setting up secure interim storage for the fissile materials and establishing an international monitoring system to verify the stockpiles and ensure that materials are not withdrawn for use in new weapons. The panel also urges Russia to stop producing fissile weapons materials and both countries to commit a very large fraction of their plutonium and highly enriched uranium from dismantled weapons to nonaggressive uses. The US and Russia have already made initial moves to accomplish these goals but have not fully implemented any of them

  12. Assessment of the environmental impact of plutonium transport within the European Community

    International Nuclear Information System (INIS)

    1978-11-01

    Assessment is given of the environmental impact of plutonium transport within the European Community in the 1990's. This includes shipments of plutonium, raw materials, fresh and spent mixed oxide fuel assemblies and plutonium contaminated solid wastes. The consequences on general traffic, the radiation doses to transport workers and the general public and the heat release during transport are presented

  13. Preparation of fused chloride salts for use in pyrochemical plutonium recovery operations at Los Alamos

    Energy Technology Data Exchange (ETDEWEB)

    Fife, K.W.; Bowersox, D.F.; Christensen, D.C.; Williams, J.D.

    1986-07-01

    The Plutonium Metal Technology Group at Los Alamos routinely uses pyrochemical processes to produce and purify plutonium from impure sources. The basic processes (metal production, metal purification, and residue treatment) involve controlling oxidation and reduction reactions between plutonium and its compounds in molten salts. Current production methods are described, as well as traditional approaches and recent developments in the preparation of solvent salts for electrorefining, molten salt extraction, lean metal (pyroredox) purification, and direct oxide reduction.

  14. Preparation of fused chloride salts for use in pyrochemical plutonium recovery operations at Los Alamos

    International Nuclear Information System (INIS)

    Fife, K.W.; Bowersox, D.F.; Christensen, D.C.; Williams, J.D.

    1986-07-01

    The Plutonium Metal Technology Group at Los Alamos routinely uses pyrochemical processes to produce and purify plutonium from impure sources. The basic processes (metal production, metal purification, and residue treatment) involve controlling oxidation and reduction reactions between plutonium and its compounds in molten salts. Current production methods are described, as well as traditional approaches and recent developments in the preparation of solvent salts for electrorefining, molten salt extraction, lean metal (pyroredox) purification, and direct oxide reduction

  15. Plutonium sorption by the green algae Scenedesmus obliquus (Tuerp) Kuetz

    International Nuclear Information System (INIS)

    Tkacik, M.F.; Giesy, J.P.; Corey, J.C.

    1978-01-01

    As part of the continuing study of the possible impact of released radioisotopes to the Savannah River Plant (SRP) environment, the interaction between a biological system and plutonium was investigated. Specifically, an algal culture, Scenedesmus obliquus, was exposed to the +4 and +6 oxidation states of 238 Pu and 239-240 Pu at three plutonium concentration levels. There was no significant different (p 3) 0.05) between 238 Pu and 239-240 Pu accumulation by the algae at equivalent concentrations or at different oxidation states

  16. Safeguards and security requirements for weapons plutonium disposition in light water reactors

    International Nuclear Information System (INIS)

    Thomas, L.L.; Strait, R.S.

    1994-10-01

    This paper explores the issues surrounding the safeguarding of the plutonium disposition process in support of the United States nuclear weapons dismantlement program. It focuses on the disposition of the plutonium by burning mixed oxide fuel in light water reactors (LWR) and addresses physical protection, material control and accountability, personnel security and international safeguards. The S and S system needs to meet the requirements of the DOE Orders, NRC Regulations and international safeguards agreements. Experience has shown that incorporating S and S measures into early facility designs and integrating them into operations provides S and S that is more effective, more economical, and less intrusive. The plutonium disposition safeguards requirements with which the US has the least experience are the implementation of international safeguards on plutonium metal; the large scale commercialization of the mixed oxide fuel fabrication; and the transportation to and loading in the LWRs of fresh mixed oxide fuel. It is in these areas where the effort needs to be concentrated if the US is to develop safeguards and security systems that are effective and efficient

  17. Effects of cocoa powder and dark chocolate on LDL oxidative susceptibility and prostaglandin concentrations in humans.

    Science.gov (United States)

    Wan, Y; Vinson, J A; Etherton, T D; Proch, J; Lazarus, S A; Kris-Etherton, P M

    2001-11-01

    Flavonoids are polyphenolic compounds of plant origin with antioxidant effects. Flavonoids inhibit LDL oxidation and reduce thrombotic tendency in vitro. Little is known about how cocoa powder and dark chocolate, rich sources of polyphenols, affect these cardiovascular disease risk factors. We evaluated the effects of a diet high in cocoa powder and dark chocolate (CP-DC diet) on LDL oxidative susceptibility, serum total antioxidant capacity, and urinary prostaglandin concentrations. We conducted a randomized, 2-period, crossover study in 23 healthy subjects fed 2 diets: an average American diet (AAD) controlled for fiber, caffeine, and theobromine and an AAD supplemented with 22 g cocoa powder and 16 g dark chocolate (CP-DC diet), providing approximately 466 mg procyanidins/d. LDL oxidation lag time was approximately 8% greater (P = 0.01) after the CP-DC diet than after the AAD. Serum total antioxidant capacity measured by oxygen radical absorbance capacity was approximately 4% greater (P = 0.04) after the CP-DC diet than after the AAD and was positively correlated with LDL oxidation lag time (r = 0.32, P = 0.03). HDL cholesterol was 4% greater after the CP-DC diet (P = 0.02) than after the AAD; however, LDL-HDL ratios were not significantly different. Twenty-four-hour urinary excretion of thromboxane B(2) and 6-keto-prostaglandin F(1)(alpha) and the ratio of the 2 compounds were not significantly different between the 2 diets. Cocoa powder and dark chocolate may favorably affect cardiovascular disease risk status by modestly reducing LDL oxidation susceptibility, increasing serum total antioxidant capacity and HDL-cholesterol concentrations, and not adversely affecting prostaglandins.

  18. U.S. weapons-usable plutonium disposition policy: Implementation of the MOX fuel option

    Energy Technology Data Exchange (ETDEWEB)

    Woods, A.L. [ed.] [Amarillo National Resource Center for Plutonium, TX (United States); Gonzalez, V.L. [Texas A and M Univ., College Station, TX (United States). Dept. of Political Science

    1998-10-01

    A comprehensive case study was conducted on the policy problem of disposing of US weapons-grade plutonium, which has been declared surplus to strategic defense needs. Specifically, implementation of the mixed-oxide fuel disposition option was examined in the context of national and international nonproliferation policy, and in contrast to US plutonium policy. The study reveals numerous difficulties in achieving effective implementation of the mixed-oxide fuel option including unresolved licensing and regulatory issues, technological uncertainties, public opposition, potentially conflicting federal policies, and the need for international assurances of reciprocal plutonium disposition activities. It is believed that these difficulties can be resolved in time so that the implementation of the mixed-oxide fuel option can eventually be effective in accomplishing its policy objective.

  19. U.S. weapons-useable plutonium disposition policy: Implementation of the MOX fuel option

    International Nuclear Information System (INIS)

    Woods, A.L.; Gonzalez, V.L.

    1998-10-01

    A comprehensive case study was conducted on the policy problem of disposing of US weapons-grade plutonium, which has been declared surplus to strategic defense needs. Specifically, implementation of the mixed-oxide fuel disposition option was examined in the context of national and international nonproliferation policy, and in contrast to US plutonium policy. The study reveals numerous difficulties in achieving effective implementation of the mixed-oxide fuel option including unresolved licensing and regulatory issues, technological uncertainties, public opposition, potentially conflicting federal policies, and the need for international assurances of reciprocal plutonium disposition activities. It is believed that these difficulties can be resolved in time so that the implementation of the mixed-oxide fuel option can eventually be effective in accomplishing its policy objective

  20. A Mixed-Oxide Assembly Design for Rapid Disposition of Weapons Plutonium in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Adams, Marvin L.

    2002-01-01

    We have created a new mixed-oxide (MOX) fuel assembly design for standard pressurized water reactors (PWRs). Design goals were to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which we call MIX-33, offers some advantages for the disposition of weapons-grade plutonium; it increases the disposition rate by 8% while increasing the worth of control material, compared to a previous Westinghouse design. The MIX-33 design is based upon two ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the addition of water holes in the assembly. The main result of this paper is that both of these ideas are effective at increasing Pu throughput and increasing the worth of control material. With this new design, according to our analyses, we can transition smoothly from a full low-enriched-uranium (LEU) core to a full MIX-33 core while meeting the operational and safety requirements of a standard PWR. Given an interruption of the MOX supply, we can transition smoothly back to full LEU while meeting safety margins and using standard LEU assemblies with uniform pinwise enrichment distribution. If the MOX supply is interrupted for only one cycle, the transition back to a full MIX-33 core is not as smooth; high peaking could cause power to be derated by a few percent for a few weeks at the beginning of one transition cycle

  1. Studies of the conversion-chemistry of plutonium and uranium in the nitrate- and carbonate-systems

    International Nuclear Information System (INIS)

    Hoffmann, G.; Steinhauser, M.; Boehm, M.

    1988-01-01

    A novel type construction of an autoclave for dissolving of plutonium dioxide in concentrated nitric acid (without any admixtures) has been developed. This process allows the dissolving of batches with high oxide/acid ratio and yields plutonium-solutions of high concentration. The tests for separation of plutonium- and, respectively, uranium-process-solutions from Am-241 and other interfering impurities are described. The time-factor for the oxidation-reaction of plutonium in nitric acid with ozone has been optimized. Important data on the solubility-behavior of plutonyl(VI)- and of pure Pu(IV)-nitrates have been gained. The majority of the precipitates, occuring in theses reactions, were characterized. (orig.) [de

  2. Experience with thermal recycle of plutonium and uranium

    International Nuclear Information System (INIS)

    Beer, O.; Schlosser, G.; Spielvogel, F.

    1985-01-01

    The Federal Republic of Germany (FRG) decided to close the fuel cycle by erecting the reprocessing plant WA350 at Wackersdorf. As long as the plutonium supply from reprocessing plants exceeds the plutonium demand of fast breeder reactors, recycling of plutonium in LWR's is a convenient solution by which a significant advanced uranium utilization is achieved. The demonstration of plutonium recycling performed to date in the FRG in BWR's and PWR's shows that thermal plutonium recycling on an industrial scale is feasible and that the usual levels of reliability and safety can be achieved in reactor operation. The recycling of reprocessed uranium is presently demonstrated in the FRG, too. As regards fuel cycle economy thermal recycling allows savings in natural uranium and separative work. Already under present cost conditions the fuel cycle costs for mixed oxide or enriched reprocessed uranium fuel assemblies are equal or even lower than for usual uranium fuel assemblies

  3. Plutonium-enriched thermal fuel production experience in Belgium

    International Nuclear Information System (INIS)

    LeBlanc, J.M.

    1983-01-01

    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

  4. Synthesis and Characterization of Oxide Feedstock Powders for the Fuel Cycle R and D Program

    International Nuclear Information System (INIS)

    Voit, Stewart L.; Vedder, Raymond James; Johnson, Jared A.

    2010-01-01

    Nuclear fuel feedstock properties, such as physical, chemical, and isotopic characteristics, have a significant impact on the fuel fabrication process and, by extension, the in-reactor fuel performance. This has been demonstrated through studies with UO 2 spanning greater than 50 years. The Fuel Cycle R and D Program with The Department of Energy Office of Nuclear Energy has initiated an effort to develop a better understanding of the relationships between oxide feedstock, fresh fuel properties, and in-reactor fuel performance for advanced mixed oxide compositions. Powder conditioning studies to enable the use of less than ideal powders for ceramic fuel pellet processing are ongoing at Los Alamos National Laboratory (LANL) and an understanding of methods to increase the green density and homogeneity of pressed pellets has been gained for certain powders. Furthermore, Oak Ridge National Laboratory (ORNL) is developing methods for the co-conversion of mixed oxides along with techniques to analyze the degree of mixing. Experience with the fabrication of fuel pellets using co-synthesized multi-constituent materials is limited. In instances where atomically mixed solid solutions of two or more species are needed, traditional ceramic processing methods have been employed. Solution-based processes may be considered viable synthesis options, including co-precipitation (AUPuC), direct precipitation, direct-conversion (Modified Direct Denitration or MDD) and internal/external gelation (sol-gel). Each of these techniques has various advantages and disadvantages. The Fiscal Year 2010 feedstock development work at ORNL focused on the synthesis and characterization of one batch of UO x and one batch of U 80 Ce 20 O x . Oxide material synthesized at ORNL is being shipped to LANL for fuel fabrication process development studies. The feedstock preparation was performed using the MDD process which utilizes a rotary kiln to continuously thermally denitrate double salts of ammonium

  5. Analysis of plutonium isotope ratios including 238Pu/239Pu in individual U-Pu mixed oxide particles by means of a combination of alpha spectrometry and ICP-MS.

    Science.gov (United States)

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

    2017-04-01

    Isotope ratio analysis of individual uranium-plutonium (U-Pu) mixed oxide particles contained within environmental samples taken from nuclear facilities is proving to be increasingly important in the field of nuclear safeguards. However, isobaric interferences, such as 238 U with 238 Pu and 241 Am with 241 Pu, make it difficult to determine plutonium isotope ratios in mass spectrometric measurements. In the present study, the isotope ratios of 238 Pu/ 239 Pu, 240 Pu/ 239 Pu, 241 Pu/ 239 Pu, and 242 Pu/ 239 Pu were measured for individual Pu and U-Pu mixed oxide particles by a combination of alpha spectrometry and inductively coupled plasma mass spectrometry (ICP-MS). As a consequence, we were able to determine the 240 Pu/ 239 Pu, 241 Pu/ 239 Pu, and 242 Pu/ 239 Pu isotope ratios with ICP-MS after particle dissolution and chemical separation of plutonium with UTEVA resins. Furthermore, 238 Pu/ 239 Pu isotope ratios were able to be calculated by using both the 238 Pu/( 239 Pu+ 240 Pu) activity ratios that had been measured through alpha spectrometry and the 240 Pu/ 239 Pu isotope ratios determined through ICP-MS. Therefore, the combined use of alpha spectrometry and ICP-MS is useful in determining plutonium isotope ratios, including 238 Pu/ 239 Pu, in individual U-Pu mixed oxide particles. Copyright © 2016 Elsevier B.V. All rights reserved.

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

    International Nuclear Information System (INIS)

    Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

    1999-01-01

    The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

  7. Prospects for the establishment of plutonium recycle in thermal reactors in the Foratom countries. Status and assessment

    International Nuclear Information System (INIS)

    Chamberlain, A.; Melches, C.

    1977-01-01

    The paper reviews the technical status of plutonium recycle in thermal reactors in the Foratom countries and assesses the prospect for it becoming established in the future with the implicit assumptions that uranium oxide reprocessing capacity will be installed commensurate with the projected programmes for thermal reactor installation and that there will be no insuperable environmental, security or safeguards obstacles to the use of plutonium as a fuel. It is argued that the feasibility of using plutonium as an alternative to 235 U as the fuel for thermal reactors, particularly LWRs, has been extensively demonstrated by a number of Foratom countries and the main problem areas are fuel fabrication and fuel reprocessing. Mixed-oxide fuel fabrication has been well established on the prototype plant scale using low-irradiation plutonium, but it is recognized that the future design of production-scale plants will need to cater for the significantly higher radiation levels from high burnup plutonium and meet stricter environmental requirements on operator dosage and waste arisings. The main constraint on the establishment of recycle up to now has been the lack of available plutonium owing to the absence of significant uranium-oxide fuel reprocessing capacity. An assessment of the plutonium arisings in Europe, based on the projected uranium-oxide reprocessing capacity, shows that by 1990 plutonium, surplus to FBR requirements, should be accumulating by about 10t/a, sufficient to fuel about 8000MW(e) of LWRs. A further constraint would then be the availability and technical problems of mixed-oxide reprocessing, which is one of the areas identified for international collaboration. It is concluded that whilst there is unlikely to be substantial recycle of plutonium in thermal reactors in the Foratom countries before the early 1990s, an incentive could possibly arise about that time. The strength of this incentive will depend on a number of factors including the status of

  8. Annihilation characteristics of positrons in oxide powders in relation to catalytic activities

    International Nuclear Information System (INIS)

    Ito, K.; Ohtsu, Y.; Tanigawa, S.; Enomura, A.; Tsuda, N.

    1982-01-01

    The annihilation chaaracteristics in magnesium oxide powders were studied by the measurements of Doppler broadening of annihilation radiations. MgO powders are well known as a solid base and are utilized as a catalyst for the reactions which start by extracting protons from molecules such as decomposition of alcohol. The isochronal annealing behavior of annihilation characteristics in the process Mg(OH) 2 → MgO was found to correspond just to the change in the number of basic points, specific surface area and catalytic activities in some reactions. From the results of the thermal equilibrium measurements of MgO powders after dehydration, the temperature dependence of S parameter can be considered as the thermal activation process of the escape of positrons from trapped states at surface to form positroniums. The derived value of this activation energy was 0.187 eV. (Auth.)

  9. Technique of preparation of plutonium screens for soft x ray spectrography (1963); Technique de preparation d'ecrans de plutonium pour la spectrographie de rayons x mous (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-07-01

    The present work concerns the preparation of thin layers of pure plutonium (thickness 100 to 10000 A) by thermal vacuum evaporation. The protection of the plutonium against oxidation is obtained by vacuum deposition of aluminium layers under. and above the plutonium layer. The purity of the layers is checked by electron and X ray diffraction which has shown that very thin films of plutonium condense in {beta} form instead of {alpha}. (author) [French] Le present travail concerne la preparation de couches minces (epaisseurs de 100 a 10000 A) de plutonium pur par evaporation thermique sous vide. La protection du plutonium contre l'oxydation par l'air est obtenu grace a des couches d'aluminium deposees sous vide sous et sur la couche de plutonium. La purete des couches est verifiee par diffraction d'electrons et de rayons X ce qui a permis d'observer que les couches tres minces de plutonium se condensent en phase {beta} au lieu de la phase {alpha}. (auteur)

  10. Low power loss and field-insensitive permeability of Fe-6.5%Si powder cores with manganese oxide-coated particles

    Energy Technology Data Exchange (ETDEWEB)

    Li, Junnan, E-mail: junnanli1991@163.com, E-mail: rzhgong@hust.edu.cn; Wang, Xian; Xu, Xiaojun; Gong, Rongzhou, E-mail: junnanli1991@163.com, E-mail: rzhgong@hust.edu.cn; Feng, Zekun [School of Optical and Electronic Information, Huazhong University of Science and Technology, Wuhan 430074 (China); Chen, Yajie; Harris, V. G. [Department of Electrical and Computer Engineering, Center for Microwave Magnetic Materials and Integrated Circuits, Northeastern University, Boston, Massachusetts 02115 (United States)

    2015-05-07

    Fe-6.5%Si alloy powders coated with manganese oxides using an innovative in situ process were investigated. The in-situ coating of the insulating oxides was realized with a KMnO{sub 4} solution by a chemical process. The insulating manganese oxides with mixed valance state were verified by X-ray photoelectron spectroscopy analysis. The thickness of the insulating layer on alloy particles was determined to be in a range of 20–210 nm, depending upon the KMnO{sub 4} concentration. The powder core loss and the change in permeability under a DC-bias field were measured at frequencies ranging from 50 to 100 kHz. The experiments indicated that the Fe-6.5%Si powder cores with a 210 nm-thick manganese oxide layer not only showed a low core loss of 459 mW/cm{sup 3} at 100 kHz but also showed a small reduction in permeability (μ(H)/μ(0) = 85% for μ = 42) at a DC-bias field of 80 Oe. This work has defined a novel pathway to realizing low core loss and field-insensitive permeability for Fe-Si powder cores.

  11. Non-proliferation issues with weapons-usable plutonium

    International Nuclear Information System (INIS)

    Gray, L.W.

    2000-01-01

    In this paper author deals with the plutonium produced in power reactors and with their using. Excess plutonium, mineralized in a ceramic matrix and incised in HLW glass, is a less attractive target for terrorist groups than either aged, irradiated weapons grade MOX fuel, or aged, U oxide spent fuel. This is especially true after the Russian and United States' Pu Disposition Programs have been completed, until the material (spent MOX fuel or the immobilized form) is stored in a sealed, repository. (authors)

  12. Plutonium stabilization and packaging system

    International Nuclear Information System (INIS)

    1996-01-01

    This document describes the functional design of the Plutonium Stabilization and Packaging System (Pu SPS). The objective of this system is to stabilize and package plutonium metals and oxides of greater than 50% wt, as well as other selected isotopes, in accordance with the requirements of the DOE standard for safe storage of these materials for 50 years. This system will support completion of stabilization and packaging campaigns of the inventory at a number of affected sites before the year 2002. The package will be standard for all sites and will provide a minimum of two uncontaminated, organics free confinement barriers for the packaged material

  13. Investigation into the feasibility of alternative plutonium shipping forms

    International Nuclear Information System (INIS)

    Mishima, J.; Lindsey, C.G.

    1983-06-01

    Pacific Northwest Laboratory (PNL), operated for the Department of Energy by the Battelle Memorial Institute, is conducting a study for the Nuclear Regulatory Commission on the feasibility of altering current plutonium shipping forms to reduce or eliminate the airborne dispersibility of PuO 2 which might occur during a shipping accident. Plutonium used for fuel fabrication is currently shipped as a PuO 2 powder with a significant fraction in the respirable size range. If the high-strength container is breached due to stresses imposed during a transportation accident, the PuO 2 powder could be subject to airborne dispersion. The available information indicated that a potential accident involving fire accompanied by crush/impact forces would lead to failure of current surface shipping containers (no assumptions were made on the possibility of such a severe accident). Criteria were defined for an alternate shipping form to mitigate the effects of such an accident. Candidate techniques and materials were evaluated as alternate shipping forms by a task team consisting of personnel from PNL and Rockwell Hanford Operations (RHO). At this time, the most promising candidate for an alternate plutonium shipping form appears to be pressing PuO 2 into unsintered (green) pellets. These green pellets satisfy the criteria for a less dispersible form without requiring significant process changes. Discussions of all candidates considered are contained in a series of appendices. Recommendations for further investigations of the applicability of green pellets as an alternate shipping form are given, including the need for a cost-benefit study

  14. Base case Pu-nitrate to Pu-oxide conversion plant

    International Nuclear Information System (INIS)

    1978-10-01

    This paper explains that the plutonium recovered in the course of reprocessing spent fuel is obtained in the form of plutonium nitrate, whereas PuO 2 is required for the fabrication of fuel rods. The reference conversion plant described in the paper converts plutonium nitrate into plutonium dioxide powder by precipitation with oxalic acid followed by calcination. The paper also describes the main features of the safety, maintenance and safeguards philosophy used in its design

  15. Chemical characterization of local and stratospheric plutonium in Ohio soils

    International Nuclear Information System (INIS)

    Muller, R.N.

    1978-01-01

    The chemical nature of plutonium derived from stratospheric fallout and industrial sources was studied in three agricultural soils. The majority of the soil plutonium was associated with a reductant-soluble, hydrous oxide phase that, under most conditions of terrestrial ecosystems, remains essentially immobile. The proportion of plutonium associated with organic matter (0.1N NaOH-extractable) varied among soils, and increased with decreasing particle size in the same soil. In a soil containing 238 Pu from a local fabrication facility and 239 , 240 Pu from stratospheric fallout, isotopic ratios between the NaOH-extractable and residual phases were essentially constant, indicating that, in these soils, plutonium from both sources behaves similarly. The distribution of soil plutonium with particle size appears to be most directly related to the mass of the soil particle

  16. Plutonium in nature; Le plutonium dans la nature

    Energy Technology Data Exchange (ETDEWEB)

    Madic, C.

    1994-12-31

    Plutonium in nature comes from natural sources and anthropogenic ones. Plutonium at the earth surface comes principally from anthropogenic sources. It is easily detectable in environment. The plutonium behaviour in environment is complex. It seems necessary for the future to reduce releases in environment, to improve predictive models of plutonium behaviour in geosphere, to precise biological impact of anthropogenic plutonium releases.

  17. Simulated physical inventory verification exercise at a mixed-oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Reilly, D.; Augustson, R.

    1985-01-01

    A physical inventory verification (PIV) was simulated at a mixed-oxide fuel fabrication facility. Safeguards inspectors from the International Atomic Energy Agency (IAEA) conducted the PIV exercise to test inspection procedures under ''realistic but relaxed'' conditions. Nondestructive assay instrumentation was used to verify the plutonium content of samples covering the range of material types from input powders to final fuel assemblies. This paper describes the activities included in the exercise and discusses the results obtained. 5 refs., 1 fig., 6 tabs

  18. Time and Temperature Test Results for PFP Thermal Stabilization Furnaces

    International Nuclear Information System (INIS)

    COMPTON, J.A.

    2000-01-01

    The national standard for plutonium storage acceptability (standard DOE-STD-3013-99, generally known as ''the 3013 standard'') has been revised to clarify the requirement for processes that will produce acceptable storage materials. The 3013 standard (Reference 1) now states that ''Oxides shall be stabilized by heating the material in an oxidizing atmosphere to a Material Temperature of at least 950 C (1742 F) for not less than 2 hours.'' The process currently in use for producing stable oxides for storage at the Plutonium Finishing Plant (PFP) heats a furnace atmosphere to 1000 C and holds it there for 2 hours. The temperature of the material being stabilized is not measured directly during this process. The Plutonium Process Support Laboratories (PPSL) were requested to demonstrate that the process currently in use at PFP is an acceptable method of producing stable plutonium dioxide consistently. A spare furnace identical to the production furnaces was set up and tested under varying conditions with non-radioactive surrogate materials. Reference 2 was issued to guide the testing program. The process currently in use at the PFP for stabilizing plutonium-bearing powders was shown to heat all the material in the furnace to at least 950 C for at least 2 hours. The current process will work for (1) relatively pure plutonium dioxide, (2) dioxide powders mixed with up to 20 weight percent magnesium oxide, and (3) dioxide powders with up to 11 weight percent magnesium oxide and 20 weight percent magnesium nitrate hexahydrate. Time and temperature data were also consistent with a successful demonstration for a mixture containing 10 weight percent each of sodium and potassium chloride; however, the molten chloride salts destroyed the thermocouples in the powder and temperature data were unavailable for part of that run. These results assume that the current operating limits of no more than 2500 grams per furnace charge and a powder height of no more than 1.5 inches remain

  19. Design aspects of plutonium air-transportable packages

    International Nuclear Information System (INIS)

    Allen, G.C.; Moya, J.L.; Pierce, J.D.; Attaway, S.W.

    1989-01-01

    Recent worldwide interest in transporting plutonium powders by air has created a need for expanding the packaging technology base as well as improving their understanding of how plutonium air transport (PAT) packagings perform during severe accident tests. Historically it has not been possible to establish design rules for individual package components because of the complex way parts interacted in forming a successful whole unit. Also, computer analyses were only considered valid for very limited portions of the design effort because of large deformations, localized tearing occurring in the package during accident testing, and extensive use of orthotropic materials. Consequently, iterative design and experimentation has historically been used to develop plutonium air-transportable packages. Full-scale prototypes have been tested since scaling of packages utilizing wood as an energy absorber and thermal insulator has not proven to be very successful. This is because the wood grain and dynamic performance of the wood during crush do not always scale. The high cost of full-scale testing of large packages has certainly hindered obtaining additional data and development new designs. The testing criteria for PAT packages, as described in the US Nuclear Regulatory Commission's Qualification Criteria to Certify a Package for Air Transport of Plutonium, NUREG-0360, 1978, are summarized. Computer modeling techniques have greatly improved over the last ten years, and there are some areas of opportunity for future applications to plutonium air-transportable package design problems. Having developed a better understanding of the performance of current packages, they have the opportunity to make major improvements in new packaging concepts. Each of these areas is explored in further depth to establish their impact on design practices for air-transportable packages

  20. Separation of trace uranium from plutonium for subsequent analysis

    International Nuclear Information System (INIS)

    Marsh, S.F.

    1980-08-01

    Trace uranium quantities are separated from plutonium metal and plutonium oxide for subsequent analysis. Samples are dissolved in hydrobromic acid or a hydrobromic acid-hydrofluoric acid mixture. The U(VI)-halide complex is separated from nonsorbed Pu(III) on an anion exchange column using sequential washes of 9M HBr, a 0.1M HI-12M HCl mixture and 0.1M HCl

  1. Chemical species of plutonium in Hanford radioactive tank waste

    International Nuclear Information System (INIS)

    Barney, G.S.

    1997-01-01

    Large quantities of radioactive wastes have been generated at the Hanford Site over its operating life. The wastes with the highest activities are stored underground in 177 large (mostly one million gallon volume) concrete tanks with steel liners. The wastes contain processing chemicals, cladding chemicals, fission products, and actinides that were neutralized to a basic pH before addition to the tanks to prevent corrosion of the steel liners. Because the mission of the Hanford Site was to provide plutonium for defense purposes, the amount of plutonium lost to the wastes was relatively small. The best estimate of the amount of plutonium lost to all the waste tanks is about 500 kg. Given uncertainties in the measurements, some estimates are as high as 1,000 kg (Roetman et al. 1994). The wastes generally consist of (1) a sludge layer generated by precipitation of dissolved metals from aqueous wastes solutions during neutralization with sodium hydroxide, (2) a salt cake layer formed by crystallization of salts after evaporation of the supernate solution, and (3) an aqueous supernate solution that exists as a separate layer or as liquid contained in cavities between sludge or salt cake particles. The identity of chemical species of plutonium in these wastes will allow a better understanding of the behavior of the plutonium during storage in tanks, retrieval of the wastes, and processing of the wastes. Plutonium chemistry in the wastes is important to criticality and environmental concerns, and in processing the wastes for final disposal. Plutonium has been found to exist mainly in the sludge layers of the tanks along with other precipitated metal hydrous oxides. This is expected due to its low solubility in basic aqueous solutions. Tank supernate solutions do not contain high concentrations of plutonium even though some tanks contain high concentrations of complexing agents. The solutions also contain significant concentrations of hydroxide which competes with other

  2. Stabilization of Rocky Flats combustible residues contaminated with plutonium metal and organic solvents

    International Nuclear Information System (INIS)

    Bowen, S.M.; Cisneros, M.R.; Jacobson, L.L.; Schroeder, N.C.; Ames, R.L.

    1998-01-01

    This report describes tests on a proposed flowsheet designed to stabilize combustible residues that were generated at the Rocky Flats Environmental Technology Site (RFETS) during the machining of plutonium metal. Combustible residues are essentially laboratory trash contaminated with halogenated organic solvents and plutonium metal. The proposed flowsheet, designed by RFETS, follows a glovebox procedure that includes (1) the sorting and shredding of materials, (2) a low temperature thermal desorption of solvents from the combustible materials, (3) an oxidation of plutonium metal with steam, and (4) packaging of the stabilized residues. The role of Los Alamos National Laboratory (LANL) in this study was to determine parameters for the low temperature thermal desorption and steam oxidation steps. Thermal desorption of carbon tetrachloride (CCl 4 ) was examined using a heated air stream on a Rocky Flats combustible residue surrogate contaminated with CCl 4 . Three types of plutonium metal were oxidized with steam in a LANL glovebox to determine the effectiveness of this procedure for residue stabilization. The results from these LANL experiments are used to recommend parameters for the proposed RFETS stabilization flowsheet

  3. General consideration of effective plutonium utilization in future LWRs

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Okubo, Tsutomu

    2009-01-01

    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)

  4. Computation Results from a Parametric Study to Determine Bounding Critical Systems of Homogeneously Water-Moderated Mixed Plutonium--Uranium Oxides

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Y.

    2001-01-11

    This report provides computational results of an extensive study to examine the following: (1) infinite media neutron-multiplication factors; (2) material bucklings; (3) bounding infinite media critical concentrations; (4) bounding finite critical dimensions of water-reflected and homogeneously water-moderated one-dimensional systems (i.e., spheres, cylinders of infinite length, and slabs that are infinite in two dimensions) that were comprised of various proportions and densities of plutonium oxides and uranium oxides, each having various isotopic compositions; and (5) sensitivity coefficients of delta k-eff with respect to critical geometry delta dimensions were determined for each of the three geometries that were studied. The study was undertaken to support the development of a standard that is sponsored by the International Standards Organization (ISO) under Technical Committee 85, Nuclear Energy (TC 85)--Subcommittee 5, Nuclear Fuel Technology (SC 5)--Working Group 8, Standardization of Calculations, Procedures and Practices Related to Criticality Safety (WG 8). The designation and title of the ISO TC 85/SC 5/WG 8 standard working draft is WD 14941, ''Nuclear energy--Fissile materials--Nuclear criticality control and safety of plutonium-uranium oxide fuel mixtures outside of reactors.'' Various ISO member participants performed similar computational studies using their indigenous computational codes to provide comparative results for analysis in the development of the standard.

  5. Plutonium Finishing Plant safety evaluation report

    International Nuclear Information System (INIS)

    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

  6. Stabilization and immobilization of military plutonium: A non-proliferation perspective

    Energy Technology Data Exchange (ETDEWEB)

    Leventhal, P. [Nuclear Control Institute, Washington, DC (United States)

    1996-05-01

    The Nuclear Control Institute welcomes this DOE-sponsored technical workshop on stabilization and immobilization of weapons plutonium (W Pu) because of the significant contribution it can make toward the ultimate non-proliferation objective of eliminating weapons-usable nuclear material, plutonium and highly enriched uranium (HEU), from world commerce. The risk of theft or diversion of these materials warrants concern, as only a few kilograms in the hands of terrorists or threshold states would give them the capability to build nuclear weapons. Military plutonium disposition questions cannot be addressed in isolation from civilian plutonium issues. The National Academy of Sciences has urged that {open_quotes}further steps should be taken to reduce the proliferation risks posed by all of the world`s plutonium stocks, military and civilian, separated and unseparated...{close_quotes}. This report discusses vitrification and a mixed oxide fuels option, and the effects of disposition choices on civilian plutonium fuel cycles.

  7. Preparation of Phosphonic Acid Functionalized Graphene Oxide-modified Aluminum Powder with Enhanced Anticorrosive Properties

    Science.gov (United States)

    He, Lihua; Zhao, Yan; Xing, Liying; Liu, Pinggui; Wang, Zhiyong; Zhang, Youwei; Liu, Xiaofang

    2017-07-01

    To improve the anticorrosive performance of aluminum powder, a common functional filler in polymer coatings, we report a novel method to prepare graphene oxide modified aluminum powder (GO-Al) using 3-aminoproplyphosphoic acid as "link" agent. The GO nanosheets were firstly functionalized with 3-aminoproplyphosphoic acid (APSA) by the reaction of amine groups of APSA and the epoxy groups of GO. Subsequently, a layer of GO nanosheets uniformly and tightly covered the surface of flaky aluminum particle though the strong linking strength between -PO(OH)2 functional groups of the modified GO and aluminum. The hydrogen evolution experiment suggests that the GO attached on the aluminum powder could effectively improve the anticorrosive performance of the pigments.

  8. Standard test method for determining plutonium by controlled-potential coulometry in H2SO4 at a platinum working electrode

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1990-01-01

    1.1 This test method covers the determination of milligram quantities of plutonium in unirradiated uranium-plutonium mixed oxide having a U/Pu ratio range of 0.1 to 10. This test method is also applicable to plutonium metal, plutonium oxide, uranium-plutonium mixed carbide, various plutonium compounds including fluoride and chloride salts, and plutonium solutions. 1.2 The recommended amount of plutonium for each aliquant in the coulometric analysis is 5 to 10 mg. Precision worsens for lower amounts of plutonium, and elapsed time of electrolysis becomes impractical for higher amounts of plutonium. 1.3 The values stated in SI units are to be regarded as standard. No other units are to be regarded as standard. 1.4 This standard does not purport to address all of the safety concens, 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. Specific precaution...

  9. Removal of plutonium from real time waste using supercritical fluid extraction

    International Nuclear Information System (INIS)

    Sujatha, K.; Sivaraman, N.; Kumar, R.; Srinivasan, T.G.; Vasudeva Rao, P.R.

    2010-01-01

    Supercritical fluid extraction (SFE) technique was carried out for the recovery of plutonium from cellulose waste matrix using supercritical carbon dioxide (SC-CO 2 ) modified with suitable ligands such as octylphenyl N,N-diisobutyl carbamoylmethyl phosphine oxide (φCMPO), tri-n-butyl phosphate (TBP), acetyl acetone, trifluoro acetyl acetone and theonyltrifluoroacetyl acetone (TTA). The maximum plutonium recovery was found to be 99.8% when SC-CO 2 modified with CMPO was employed. About 15mg of plutonium was recovered from waste. (author)

  10. Test plan for demonstrating plutonium extraction from 10-L solutions using EIChrom extraction chromatographic resins

    International Nuclear Information System (INIS)

    Barney, G.S.

    1994-01-01

    Corrosive plutonium solutions stored in 10-L containers at the Plutonium Finishing Plant must be treated to convert the plutonium to a safe, solid form for storage and to remove the americium so that radiation exposure can be reduced. Extraction chromatographic resins will be tested for separating plutonium from these solutions in the laboratory. Separation parameters will be developed during the testing for large scale processing of the 10-L solutions and solutions of similar composition. Use of chromatographic resins will allow plutonium separation with minimum of chemical addition to the feed and without the need for plutonium valence adjustment. The separated plutonium will be calcined to plutonium oxide by direct solution calcination

  11. The transports in the French Plutonium Industry. A high risk activity

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-02-01

    This study throws light on the scale of transport of plutonium in France nuclear industry, an activity involving quantities of high risk materials often unknown to the public. The study is a significantly extended update of the one carried out by WISE-Paris in 1995 for the Plutonium Forum. It was motivated by important developments in the French plutonium industry and the publication of numerous data concerning transport activities since 1995. The 2003 study presents, in particular, all of the flows of plutonium crossing France every year, as well as analysis of the risks associated with this particular transport activity. Putting these data into perspective in terms of a rapidly and permanently changing political and industrial context, and a description of the regulatory framework within which shipments of plutonium take place, serve as a guide and source of reference to help readers better understand the issues. The importance of transport in the plutonium ''chain'', i.e. the stages corresponding to various industrial processes, is often under-estimated, even by the nuclear industry itself. Transport is, in fact, the activity which involves the greatest quantities of plutonium in the entire nuclear chain. Plutonium, produced during the fission reactions in the cores of nuclear reactors, is transported, contained in the irradiated fuel, to the facilities at La Hague where reprocessing separates it from the other radioactive components of the spent fuel. Part of the plutonium, now isolated in powder form, is then shipped to one of the three plants able to produce the fuel known as MOX. These are located at Cadarache and Marcoule, in France, and at Dessel in Belgium. Once in the MOX form, this plutonium has to be re-transported to reactor sites to be used. Once irradiated, the spent MOX will return to the La Hague installations to be stored for an unknown period; the plutonium contained in the spent MOX is not, at present, destined to be re

  12. The transports in the French Plutonium Industry. A high risk activity

    International Nuclear Information System (INIS)

    2003-02-01

    This study throws light on the scale of transport of plutonium in France nuclear industry, an activity involving quantities of high risk materials often unknown to the public. The study is a significantly extended update of the one carried out by WISE-Paris in 1995 for the Plutonium Forum. It was motivated by important developments in the French plutonium industry and the publication of numerous data concerning transport activities since 1995. The 2003 study presents, in particular, all of the flows of plutonium crossing France every year, as well as analysis of the risks associated with this particular transport activity. Putting these data into perspective in terms of a rapidly and permanently changing political and industrial context, and a description of the regulatory framework within which shipments of plutonium take place, serve as a guide and source of reference to help readers better understand the issues. The importance of transport in the plutonium ''chain'', i.e. the stages corresponding to various industrial processes, is often under-estimated, even by the nuclear industry itself. Transport is, in fact, the activity which involves the greatest quantities of plutonium in the entire nuclear chain. Plutonium, produced during the fission reactions in the cores of nuclear reactors, is transported, contained in the irradiated fuel, to the facilities at La Hague where reprocessing separates it from the other radioactive components of the spent fuel. Part of the plutonium, now isolated in powder form, is then shipped to one of the three plants able to produce the fuel known as MOX. These are located at Cadarache and Marcoule, in France, and at Dessel in Belgium. Once in the MOX form, this plutonium has to be re-transported to reactor sites to be used. Once irradiated, the spent MOX will return to the La Hague installations to be stored for an unknown period; the plutonium contained in the spent MOX is not, at present, destined to be re-used. (author)

  13. Magnetically responsive enzyme powders

    Energy Technology Data Exchange (ETDEWEB)

    Pospiskova, Kristyna, E-mail: kristyna.pospiskova@upol.cz [Regional Centre of Advanced Technologies and Materials, Palacky University, Slechtitelu 11, 783 71 Olomouc (Czech Republic); Safarik, Ivo, E-mail: ivosaf@yahoo.com [Regional Centre of Advanced Technologies and Materials, Palacky University, Slechtitelu 11, 783 71 Olomouc (Czech Republic); Department of Nanobiotechnology, Institute of Nanobiology and Structural Biology of GCRC, Na Sadkach 7, 370 05 Ceske Budejovice (Czech Republic)

    2015-04-15

    Powdered enzymes were transformed into their insoluble magnetic derivatives retaining their catalytic activity. Enzyme powders (e.g., trypsin and lipase) were suspended in various liquid media not allowing their solubilization (e.g., saturated ammonium sulfate and highly concentrated polyethylene glycol solutions, ethanol, methanol, 2-propanol) and subsequently cross-linked with glutaraldehyde. Magnetic modification was successfully performed at low temperature in a freezer (−20 °C) using magnetic iron oxides nano- and microparticles prepared by microwave-assisted synthesis from ferrous sulfate. Magnetized cross-linked enzyme powders were stable at least for two months in water suspension without leakage of fixed magnetic particles. Operational stability of magnetically responsive enzymes during eight repeated reaction cycles was generally without loss of enzyme activity. Separation of magnetically modified cross-linked powdered enzymes from reaction mixtures was significantly simplified due to their magnetic properties. - Highlights: • Cross-linked enzyme powders were prepared in various liquid media. • Insoluble enzymes were magnetized using iron oxides particles. • Magnetic iron oxides particles were prepared by microwave-assisted synthesis. • Magnetic modification was performed under low (freezing) temperature. • Cross-linked powdered trypsin and lipase can be used repeatedly for reaction.

  14. Use of sulfuric-nitric acid for the recovery of plutonium from HEPA filters. (620.2, WH001/LWE)

    International Nuclear Information System (INIS)

    Clark, D.E.

    1978-09-01

    Contaminated high-efficiency particulate air (HEPA) filter media, containing PuO 2 powder which had been calcined at 700 0 C, were treated with concentrated H 2 SO 4 -HNO 3 at 190 to 200 0 C for periods ranging from 0.5 to 2.0 hours. When followed by a dilute HNO 3 rinse, this treatment was shown to be very effective as a plutonium recovery operation (approximately greater than 97% of the plutonium was solubilized). A proposed treatment scheme is given which could provide both a plutonium recovery option for HEPA filters and a reduction in overall waste volume

  15. Development of Pneumatic Transport System (PTS) for safe handling of uranium oxide powder in UMP/UED

    International Nuclear Information System (INIS)

    Manna, S.; Satpati, S.K.; Roy, S.B.

    2009-01-01

    Tonnage quantity radioactive uranium oxide powder of particle size sub micron to 100 micron is handled in Uranium Metal Plant (UMP), UED/BARC for production of nuclear grade uranium metal, required for fuelling research reactors - Dhruva and Cirus. A Pneumatic Transfer System (PTS) using vacuum has been introduced and is being used for handling radioactive powder to improve radiation protection

  16. Analysis and characterization of plutonium in pyrochemical salt residues

    International Nuclear Information System (INIS)

    Haschke, John M.; Phillips, Alan G.

    2000-01-01

    Quantitative measurement of hydrogen produced during salt-catalyzed hydrolysis of plutonium in pyrochemical salt residues show that the metal is present in concentrations of 10 ± 5 mass%. The analytical method is based on stoichiometric reaction of metal with water to form plutonium monoxide monohydride (PuOH) and hydrogen. Results of a kinetic model developed to describe the observed time dependence of the hydrolysis rate shows that metal is present predominately as particles with essentially spherical geometries and diameters near 1 mm. The presence of smaller metallic particles cannot be verified or excluded. Plutonium concentrations measured for residues stored in different configurations suggest that a sizable fraction of the metal has oxidized during storage. Application of the hydrolysis method in determining concentrations and dimensions of metallic plutonium in other nuclear waste forms is proposed

  17. 1981 Annual Status Report. Plutonium fuels and actinide programme

    International Nuclear Information System (INIS)

    1981-01-01

    In this 1981 report the work carried out by the European Institute for Transuranium elements is reviewed. Main topics are: operation limits of plutonium fuels: swelling of advanced fuels, oxide fuel transients, equation of state of nuclear materials; actinide cycle safety: formation of actinides (FACT), safe handling of plutonium fuel (SHAPE), aspects of the head-end processing of carbide fuel (RECARB); actinide research: crystal chemistry, solid state studies, applied actinide research

  18. Effect of salt, kinnow and pomegranate fruit by-product powders on color and oxidative stability of raw ground goat meat during refrigerated storage.

    Science.gov (United States)

    Devatkal, Suresh K; Naveena, B M

    2010-06-01

    Effects of salt, kinnow and pomegranate fruit by-product powders on color and oxidative stability of raw ground goat meat stored at 4+/-1 degrees C was evaluated. Five treatments evaluated include: control (only meat), MS (meat+2% salt), KRP (meat+2% salt+2% kinnow rind powder), PRP (meat+2% salt+2% pomegranate rind powder) and PSP (meat+2% salt+2% pomegranate seed powder). Addition of salt resulted in reduction of redness scores. Lightness increased in control and unchanged in others during storage. Redness scores declined and yellowness showed inconsistent changes during storage. Thiobarbituric acid reactive substances (TBARS) values were higher (PPRP>KRP>control>MS. Therefore, these powders have potential to be used as natural antioxidants to minimize the auto-oxidation and salt induced lipid oxidation in raw ground goat meat. Crown Copyright 2010. Published by Elsevier Ltd. All rights reserved.

  19. Study of plutonium fire extinction in glove-boxes

    International Nuclear Information System (INIS)

    Boisson, H.; Chappellier, A.; Debiar, A.; Savornin, J.

    1965-01-01

    The report concerns attempts to extinguish fire in uranium - chosen as a basis for comparison - and in plutonium, these being contained in glove boxes. The extinguishing powders TEC and MG 10 were tried and compared. The operator used in plastic bags. The metal was dealt with respectively in the form of billets, compact chips and finally thin shavings. In each case fire control was without difficulty. One must allow a certain time for cooling. The powder MG 10 showed very good results. Moreover, the trials proved that densely packed shavings are a very real security factor. The interior lay-out of the glove boxes ought, in order to facilitate further manipulations, be such as to render extinction as easy as possible should the need arise. (authors) [fr

  20. Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors

    International Nuclear Information System (INIS)

    Griffin, C.W.; Rasmussen, D.E.; Lloyd, M.H.

    1983-01-01

    The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities

  1. Investigations on the oxidation of nitric acid plutonium solutions with ozone

    International Nuclear Information System (INIS)

    Boehm, M.

    1983-01-01

    The reaction of ozone with nitric acid Pu solutions was studied as a function of reaction time, acid concentration and Pu concentration. Strong nitric acid Pu solutions are important in nuclear fuel element production and reprocessing. The Pu must be converted into hexavalent Pu before precipitation from the homogeneous solution together with uranium-IV, ammonia and CO 2 in the form of ammonium uranyl/plutonyl carbonate (AUPuC). Formation of a solid phase during ozonation was observed for the first time. The proneness to solidification increases with incrasing plutonium concentrations and with decreasing acid concentrations. If the formation of a solid phase during ozonation of nitric acid Pu solutions cannot be prevented, the PU-IV oxidation process described is unsuitable for industrial purposes as Pu solutions in industrial processes have much higher concentrations than the solutions used in the present investigation. (orig./EF) [de

  2. Radiation exposure and risk estimates for inhaled airborne radioactive pollutants including hot particles. Annual report 1 July 1976--30 June 1977

    International Nuclear Information System (INIS)

    Mewhinney, J.A.

    1978-03-01

    Contents: Mixed-oxide fuel fabrication; Generation of aerosols of mixed uranium-plutonium oxides from dry powders for animal inhalation exposures; Analytical radiochemical determination of U, Pu and Am in biological samples; Physical chemical characterization of mixed uranium-plutonium oxide nuclear fuel as samples during animal inhalation exposure; Pilot studies of deposition and retention of industrial mixed-oxide aerosols in the laboratory rat; Extended radiation dose pattern studies of aerosols of mixed uranium-plutonium oxides treated at 750C inhaled by Fishcer-344 rats, beagle dogs and cynomolgus monkeys; Extended radiation dose pattern studies of aerosols of plutonium dioxide, treated at 850C and inhaled by Fischer-344 rats, beagle dogs and cynomolgus monkeys

  3. Characteristics of MOX dissolution with silver mediated electrolytic oxidation method

    Energy Technology Data Exchange (ETDEWEB)

    Umeda, Miki; Nakazaki, Masato; Kida, Takashi; Sato, Kenji; Kato, Tadahito; Kihara, Takehiro; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    MOX dissolution with silver mediated electrolytic oxidation method is to be applied to the preparation of plutonium nitrate solution to be used for criticality safety experiments at Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). Silver mediated electrolytic oxidation method uses the strong oxidisation ability of Ag(II) ion. This method is though to be effective for the dissolution of MOX, which is difficult to be dissolved with nitric acid. In this paper, the results of experiments on dissolution with 100 g of MOX are described. It was confirmed from the results that the MOX powder to be used at NUCEF was completely dissolved by silver mediated electrolytic oxidation method and that Pu(VI) ion in the obtained solution was reduced to tetravalent by means of NO{sub 2} purging. (author)

  4. The first milligrams of plutonium

    International Nuclear Information System (INIS)

    Goldschmidt, B.

    1997-01-01

    A historical review of the development of the very first quantities of plutonium produced during World War II in the United States and in Canada, as remembered by the French nuclear chemist, Mr Goldschmidt, who participated to the various programs which were involved in the development of the atomic bomb, and to the first steps of the French atomic program after the war. Mr Goldschmidt worked especially on organic solvent extraction, with the selection, in 1945, of non volatile tri glycol dichloride, and the development of the Chalk River plant. In 1949, at the Bouchet plant, his team has isolated the first milligrams of French plutonium from uranium oxide; and in 1952, the PUREX process was developed

  5. The transport of plutonium industry in France. A high risk activity; Les transports de l'industrie du plutonium en France. Une activite a haut risque. Resume de l'etude et rapport final

    Energy Technology Data Exchange (ETDEWEB)

    Coeytaux, X.; Faid, Y.B.; Hazemann, J.; Marignac, Y.; Schneider, M

    2003-02-15

    According to the estimations of the report more than 450 transport of materials including about 40 tons of plutonium circulate in France during one year. The nuclear matters transported are a source of danger. The risks in relation with the handling and transport of plutonium touch the criticality risk, its great toxicity, and the problem of proliferation. To these risks it is necessary to add terrorists attacks. Three scenarios of accidents have been developed to illustrate the risk potential. A scenario of train accident, with derailment in a tunnel then crash with an other train; a simple scenario occurring to a lorry of plutonium powder; then a scenario of an accident following a terrorist act. These scenarios should be studied by I.R.S.N and the results should be given to the public and to elected people confronted to the transport of plutonium. (N.C.)

  6. Direct conversion of plutonium-containing materials to borosilicate glass for storage or disposal

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Beahm, E.C.

    1995-01-01

    A new process, the Glass Material Oxidation and Dissolution System (GMODS), has been invented for the direct conversion of plutonium metal, scrap, and residue into borosilicate glass. The glass should be acceptable for either the long-term storage or disposition of plutonium. Conversion of plutonium from complex chemical mixtures and variable geometries into homogeneous glass (1) simplifies safeguards and security; (2) creates a stable chemical form that meets health, safety, and environmental concerns; (3) provides an easy storage form; (4) may lower storage costs; and (5) allows for future disposition options. In the GMODS process, mixtures of metals, ceramics, organics, and amorphous solids containing plutonium are fed directly into a glass melter where they are directly converted to glass. Conventional glass melters can accept materials only in oxide form; thus, it is its ability to accept materials in multiple chemical forms that makes GMODS a unique glass making process. Initial proof-of-principle experiments have converted cerium (plutonium surrogate), uranium, stainless steel, aluminum, and other materials to glass. Significant technical uncertainties remain because of the early nature of process development

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

    International Nuclear Information System (INIS)

    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

  8. METHOD OF SEPARATING URANIUM VALUES, PLUTONIUM VALUES AND FISSION PRODUCTS BY CHLORINATION

    Science.gov (United States)

    Brown, H.S.; Seaborg, G.T.

    1959-02-24

    The separation of plutonium and uranium from each other and from other substances is described. In general, the method comprises the steps of contacting the uranium with chlorine in the presence of a holdback material selected from the group consisting of lanthanum oxide and thorium oxide to form a uranium chloride higher than uranium tetrachloride, and thereafter heating the uranium chloride thus formed to a temperature at which the uranium chloride is volatilized off but below the volatilizalion temperature of plutonium chloride.

  9. Cleaning up the legacy of the cold war: plutonium oxides and the role of synchrotron radiation research

    International Nuclear Information System (INIS)

    Clark, David L.

    2015-01-01

    This presentation will cover three aspects of Los Alamos synchrotron radiation studies of plutonium oxides: (1) the use of UHV X-ray spectroscopy techniques to gain insights into the electronic structure of the actinide dioxides; (2) the results and interpretation Pu X-ray Absorption Fine Structure (XAFS) spectra of 24 samples of PuO 2+x ; and (3) the application of synchrotron radiation towards the decontamination and decommissioning of the Rocky Flats Environmental Technology Site (RFETS, the former nuclear weapons production site for the US). I will summarize a multidisciplinary effort to characterize the electronic structure and bonding of stoichiometric AnO 2 (U, Pu) using our screened hybrid DFT calculations for comparison with photoemission spectroscopy (PES) to measure the experimental density of states (DOS), and with oxygen K-edge X-ray absorption spectroscopy (XAS) to map the O 2p character in the unoccupied DOS. I will also describe the structural chemistry of laboratory prepared AnO 2+x systems (An = U, Pu; 0 ≤ x ≤ 0.25) using a combination of X-ray absorption fine structure spectroscopy (XAFS) and X-ray and neutron scattering of laboratory prepared samples. All data are consistent with mixed-valent solids where plutonium exists as a mixture of Pu(IV) and Pu(V) oxidation states, and uranium exists as a mixture of U(IV) and U(VI)

  10. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  11. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

    1994-01-01

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

  12. Dry sample storage system for an analytical laboratory supporting plutonium processing

    International Nuclear Information System (INIS)

    Treibs, H.A.; Hartenstein, S.D.; Griebenow, B.L.; Wade, M.A.

    1990-01-01

    The Special Isotope Separation (SIS) plant is designed to provide removal of undesirable isotopes in fuel grade plutonium by the atomic vapor laser isotope separation (AVLIS) process. The AVLIS process involves evaporation of plutonium metal, and passage of an intense beam of light from a laser through the plutonium vapor. The laser beam consists of several discrete wavelengths, tuned to the precise wavelength required to ionize the undesired isotopes. These ions are attracted to charged plates, leaving the bulk of the plutonium vapor enriched in the desired isotopes to be collected on a cold plate. Major portions of the process consist of pyrochemical processes, including direct reduction of the plutonium oxide feed material with calcium metal, and aqueous processes for purification of plutonium in residues. The analytical laboratory for the plant is called the Material and Process Control Laboratory (MPCL), and provides for the analysis of solid and liquid process samples

  13. Digital pile-up rejection for plutonium experiments with solution-grown stilbene

    Energy Technology Data Exchange (ETDEWEB)

    Bourne, M.M., E-mail: mmbourne@umich.edu; Clarke, S.D., E-mail: clarkesd@umich.edu; Paff, M., E-mail: mpaff@umich.edu; DiFulvio, A., E-mail: difulvio@umich.edu; Norsworthy, M., E-mail: marknors@umich.edu; Pozzi, S.A., E-mail: pozzisa@umich.edu

    2017-01-11

    A solution-grown stilbene detector was used in several experiments with plutonium samples including plutonium oxide, mixed oxide, and plutonium metal samples. Neutrons from different reactions and plutonium isotopes are accompanied by numerous gamma rays especially by the 59-keV gamma ray of {sup 241}Am. Identifying neutrons correctly is important for nuclear nonproliferation applications and makes neutron/gamma discrimination and pile-up rejection necessary. Each experimental dataset is presented with and without pile-up filtering using a previously developed algorithm. The experiments were simulated using MCNPX-PoliMi, a Monte Carlo code designed to accurately model scintillation detector response. Collision output from MCNPX-PoliMi was processed using the specialized MPPost post-processing code to convert neutron energy depositions event-by-event into light pulses. The model was compared to experimental data after pulse-shape discrimination identified waveforms as gamma ray or neutron interactions. We show that the use of the digital pile-up rejection algorithm allows for accurate neutron counting with stilbene to within 2% even when not using lead shielding.

  14. Plutonium

    International Nuclear Information System (INIS)

    Watson, G.M.

    1976-01-01

    Discovery of the neutron made it easy to create elements which do not exist in nature. One of these is plutonium, and its isotope with mass number 239 has nuclear properties which make it both a good fuel for nuclear power reactors and a good explosive for nuclear weapons. Since it was discovered during a war the latter characteristic was put to use, but it is now evident that use of plutonium in a particular kind of nuclear reactor, the fast breeder reactor, will allow the world's resources of uranium to last for millennia as a major source of energy. Plutonium is very radiotoxic, resembling radium in this respect. Therefore the widespread introduction of fast breeder reactors to meet energy demands can be contemplated only after assurances on two points; that adequate control of the radiological hazard resulting from the handling of very large amounts of plutonium can be guaranteed, and that diversion of plutonium to illicit use can be prevented. The problems exist to a lesser degree already, since all types of nuclear reactor produce some plutonium. Some plutonium has already been dispersed in the environment, the bulk of it from atmospheric tests of nuclear weapons. (author)

  15. Fuel qualification issues and strategies for reactor-based surplus plutonium disposition

    International Nuclear Information System (INIS)

    Cowell, B.S.; Copeland, G.L.; Moses, D.L.

    1997-08-01

    The Department of Energy (DOE) has proposed irradiation of mixed-oxide (MOX) fuel in existing commercial reactors as a disposition method for surplus plutonium from the weapons program. The burning of MOX fuel in reactors is supported by an extensive technology base; however, the infrastructure required to implement reactor-based plutonium disposition does not exist domestically. This report identifies and examines the actions required to qualify and license weapons-grade (WG) plutonium-based MOX fuels for use in domestic commercial light-water reactors (LWRs)

  16. Dynamic and quasi-static simulation and analysis of the plutonium oxide/metal containers subject to 30-foot dropping

    International Nuclear Information System (INIS)

    Gong, C.; Miller, R.F.

    1995-01-01

    This analysis of the plutonium oxide/metal storage containers is in support of the design and testing project The results from the dynamic analysis show some important facts that have not been considered before. The internal bagless transfer can will have higher stress than the primary container. The quasi-static analysis provides a conservative solution. In both vertical upright drop (dynamic) and inclined upside down drop (quasi-static) the containers are structurally sound

  17. Rupture of plutonium oxide storage container, March 13, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1979-05-29

    On March 13, 1979, a plutonium oxide storage can ruptured in the 303-C storage facility, which is in the 300 Area of the Hanford Site, Washington. The facility is operated by the Pacific Northwest Laboratory (PNL); three PNL staff members were performing the storage operation. No injuries to these staff members resulted from the occurrence. A Class C Investigation Committee was appointed on March 14, 1979, by the Director, PNL. Subsequently, when the loss estimates became available, the Manager, Department of Energy-Richland Operations Office (DOE-RL), appointed a Class B Investigation Committee in accordance with DOE Manual Chapter 0502. As requested by DOE-RL, the Committee investigated technical elements of the causal sequence and management systems that should have or could have prevented the occurrence. The investigation included: review of the use of the 303-C facilities and the transfer containers; interviews with the involved personnel and their managers; analysis of technical studies related to involved materials and procedures; review of safe operating procedures, radiation work procedures, and transfer requirements applicable to the occurrence; and use of the Management Oversight and Risk Tree (MORT) and the Events and Causal Factors Charting methods. 15 figs.

  18. Sequential potentiometric determination of uranium and plutonium in a single aliquot

    International Nuclear Information System (INIS)

    Rao, V.K.; Charyulu, M.M.; Natarajan, P.R.

    1983-01-01

    A method is reported for sequential potentiometric determination of uranium and plutonium present is an aliquot. Plutonium is first determined by oxidizing it to the hexavalent state with perchloric acid followed by iron(II) reduction and titration of excess ferrous iron with chromium(VI). Uranium is subsequently determined by reduction to the quadrivalent state using titanium(III) and titration with vanadium(V). The interference of plutonium and iron(II) is eliminated by the addition of a mixture containing sulfamic acid, nitric acid, and molybdenum(VI). The results of the analysis of mixture containing 3-5 mg quantities of uranium and plutonium are reliable with errors less than 0.3% and 0.2%, respectively. The application of the method for the analysis of mixtures containing various amounts of uranium and plutonium has been examined. (author)

  19. Surplus plutonium disposition draft environmental impact statement. Volume 2

    International Nuclear Information System (INIS)

    1998-07-01

    On May 22, 1997, DOE published a Notice of Intent (NOI) in the Federal Register (62 Federal Register 28009) announcing its decision to prepare an environmental impact statement (EIS) that would tier from the analysis and decisions reached in connection with the Storage and Disposition of Weapons-Usable Fissile Materials Final Programmatic EIS (Storage and Disposition PEIS). DOE's disposition strategy allows for both the immobilization of surplus plutonium and its use as mixed oxide (MOX) fuel in existing domestic, commercial reactors. The disposition of surplus plutonium would also involve disposal of the immobilized plutonium and MOX fuel (as spent nuclear fuel) in a geologic repository. The Surplus Plutonium Disposition Environmental Impact Statement analyzes alternatives that would use the immobilization approach (for some of the surplus plutonium) and the MOX fuel approach (for some of the surplus plutonium); alternatives that would immobilize all of the surplus plutonium; and the No Action Alternative. The alternatives include three disposition facilities that would be designed so that they could collectively accomplish disposition of up to 50 metric tons (55 tons) of surplus plutonium over their operating lives: (1) the pit disassembly and conversion facility would disassemble pits (a weapons component) and convert the recovered plutonium, as well as plutonium metal from other sources, into plutonium dioxide suitable for disposition; (2) the immobilization facility would include a collocated capability for converting nonpit plutonium materials into plutonium dioxide suitable for immobilization and would be located at either Hanford or SRS. DOE has identified SRS as the preferred site for an immobilization facility; (3) the MOX fuel fabrication facility would fabricate plutonium dioxide into MOX fuel. Volume 2 contains the appendices to the report and describe the following: Federal Register notices; contractor nondisclosure statement; adjunct melter

  20. Operating experience in processing of differently sourced deeply depleted uranium oxide and production of deeply depleted uranium metal ingots

    International Nuclear Information System (INIS)

    Manna, S.; Ladola, Y.S.; Sharma, S.; Chowdhury, S.; Satpati, S.K.; Roy, S.B.

    2009-01-01

    Uranium Metal Plant (UMP) of BARC had first time experience on production of three Depleted Uranium Metal (DUM) ingots of 76kg, 152kg and 163kg during March 1991. These ingots were produced by processing depleted uranyl nitrate solution produced at Plutonium Plant (PP), Trombay. In recent past Uranium Metal Plant (UMP), Uranium Extraction Division (UED), has been assigned to produce tonnage quantity of Deeply DUM (DDUM) from its oxide obtained from PP, PREFRE and RMP, BARC. This is required for shielding the high radioactive source of BHABHATRON Tele-cobalt machine, which is used for cancer therapy. The experience obtained in processing of various DDU oxides is being utilized for design of large scale DDU-metal plant under XIth plan project. The physico- chemical characteristics like morphology, density, flowability, reactivity, particle size distribution, which are having direct effect on reactivity of the powders of the DDU oxide powder, were studied and the shop-floor operational experience in processing of different oxide powder were obtained and recorded. During campaign trials utmost care was taken to standardized all operating conditions using the same equipment which are in use for natural uranium materials processing including safety aspects both with respect to radiological safety and industrial safety. Necessary attention and close monitoring were specially arranged and maintained for the safety aspects during the trial period. In-house developed pneumatic transport system was used for powder transfer and suitable dust arresting system was used for reduction of powder carry over

  1. Studies of plutonium in human tracheobronchial lymph nodes

    International Nuclear Information System (INIS)

    McInroy, J.F.; Stewart, M.W.; Moss, W.D.

    1976-01-01

    Since 1959, tissues from 70 occupationally exposed former employees of the Los Alamos Scientific Laboratory have been examined following autopsy. Exposure in most cases was to inhaled plutonium oxide aerosols. Chemical analyses of selected tissues were performed to determine the amount of plutonium retained in the body at the time of death. On the basis of the measured tissue concentrations of plutonium, extrapolations of total-body burdens were made. Thirty-three of the measured cases had plutonium depositions in the tracheobronchial lymph nodes ranging from 0.1 to 4000 dpm per gram of tissue (0.05 to 1800 pCi/g). The duration of exposures ranged from 4 to 30 years. Microscopic examination of representative sections of these lymph nodes revealed no abnormalities other than those which were directly attributable to the basic disease that caused the demise of the various persons in this study. The size distribution of plutonium particles in nodes from one individual was determined by exposing tissue sections to nuclear track film. The estimated mass median diameter of the particles was 0.3 μm, and the distribution had a geometric standard deviation of 1.6. It is estimated that 95 percent of the individual particles had corresponding plutonium concentrations between 0.001 and 0.22 pCi

  2. Influence of Diatomite and Mineral Powder on Thermal Oxidative Ageing Properties of Asphalt

    Directory of Open Access Journals (Sweden)

    Yongchun Cheng

    2015-01-01

    Full Text Available Ageing of asphalt affects the performances of asphalt pavement significantly. Therefore, effects of diatomite and mineral powder on ageing properties of asphalt were investigated systematically in order to improve the antiageing property of mixture. Thin film oven test (TFOT was used to conduct the short term ageing in laboratory. Softening points, penetrations, force ductility, low temperature creep properties, and viscosities of asphalt mastics were tested before and after TFOT, respectively. Results indicated that percent retained penetration (PRP increased with the increasing of fillers. Increment of softening point (ΔT, ductility retention rate (DRR, deformation energy ageing index (JAI, and viscosity ageing index (VAI of asphalt mastics nonlinearly decreased with the increasing of fillers. Ageing of asphalt was reduced by diatomite and mineral powder. And the antiageing effect of diatomite was better than that of mineral powder as a result of its porous structure. It is suggested that the mineral powder could be reasonably replaced by diatomite in order to reduce thermal oxidative ageing of asphalt mixture. The optimal content of diatomite 12.8% is also suggested for engineering.

  3. Stabilization of plutonium bearing residues at Lawrence Livermore National Laboratory

    International Nuclear Information System (INIS)

    Bronson, M.C.; Van Konynenburg, R.A.; Ebbinghaus, B.B.

    1995-01-01

    The US Department of Energy's (US DOE) Lawrence Livermore National Laboratory (LLNL) has plutonium holdings including metal, oxide and residue materials, all of which need stabilization of some type. Residue materials include calcined ash, calcined precipitates, pyrochemical salts, glove box sweepings, metallurgical samples, graphite, and pyrochemical ceramic crucibles. These residues are typical of residues stored throughout the US DOE plutonium sites. The stabilization process selected for each of these residues requires data on chemical impurities, physical attributes, and chemical forms of the plutonium. This paper outlines the characterization and stabilization of LLNL ash residues, pyrochemical salts, and graphite

  4. Automated amperometric plutonium assay system

    International Nuclear Information System (INIS)

    Burt, M.C.

    1985-01-01

    The amperometric titration for plutonium assay has been used in the nuclear industry for over twenty years and has been in routine use at the Hanford Engineering Development Laboratory since 1976 for the analysis of plutonium oxide and mixed oxide fuel material for the Fast Flux Test Facility. It has proven itself to be an accurate and reliable method. The method may be used as a direct end point titration or an excess of titrant may be added and a back titration performed to aid in determination of the end point. Due to the slowness of the PuVI-FeII reaction it is difficult to recognize when the end point is being approached and is very time consuming if the current is allowed to decay to the residual value after each titrant addition. For this reason the back titration in which the rapid FeII-CrVI reaction occurs is used by most laboratories. The back titration is performed by the addition of excess ferrous solution followed by two measured aliquots of standard dichromate with measurement of cell current after each addition

  5. Complexation of Plutonium (IV) With Sulfate At Variable Temperatures

    International Nuclear Information System (INIS)

    Y. Xia; J.I. Friese; D.A. Moore; P.P. Bachelor; L. Rao

    2006-01-01

    The complexation of plutonium(IV) with sulfate at variable temperatures has been investigated by solvent extraction method. A NaBrO 3 solution was used as holding oxidant to maintain the plutonium(IV) oxidation state throughout the experiments. The distribution ratio of Pu(IV) between the organic and aqueous phases was found to decrease as the concentrations of sulfate were increased. Stability constants of the 1:1 and 1:2 Pu(IV)-HSO 4 - complexes, dominant in the aqueous phase, were calculated from the effect of [HSO 4 - ] on the distribution ratio. The enthalpy and entropy of complexation were calculated from the stability constants at different temperatures using the Van't Hoff equation

  6. Cost-benefit analysis of unfired PuO2 pellets as an alternative plutonium shipping form

    International Nuclear Information System (INIS)

    Mishima, J.; Brackenbush, L.W.; Libby, R.A.; Soldat, K.L.; White, G.D.

    1983-10-01

    A limited cost-benefit evaluation was performed concerning use of unfired plutonium dioxide pellets as a shipping form. Two specific processing operations are required for this use, one to form the pellet (pelletizing) and a second to reconstitute an acceptable powder upon receipt (reconstitution). The direct costs for the pelletizing operation are approximately $208,000 for equipment and its installation and $122 per kg of plutonium processed (based upon a 20-kg plutonium/day facility). The direct costs for reconstitution are approximately $90,000 for equipment and its installation and $81 per kg of plutonium processed. The indirect cost considered was personnel exposure from these operations. Whole body exposures ranged from 0.04 man-rem per 100 kg of low-exposure plutonium reconstituted to 0.9 man-rem per 100 kg of average-exposure plutonium pelletized. Hand exposures were much higher - 17 man-rem power 100 kg of low-exposure plutonium reconstituted to 67 man-rem per 100 kg of average plutonium pelletized. The principal benefit is a potential twentyfold reduction of airborne release in the event of an accident. An experimental plan is outlined to fill the data gaps uncovered during this study in the areas of pelletizing and reconstitution process parameters and pellet response behavior to accident-generated stresses. A study to enhance the containment potential of the inner packaging used during shipment is also outlined

  7. 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)

    Vallet, V.

    2012-01-01

    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 233 U, 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 233 U, 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 ThPuO 2 fuel, and the second is with standard or high MR and ThUO 2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233 U 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 233 U 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 233 U production is the limiting factor. That is why it was eventually proposed to study how the production of 233 U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233 U 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) [fr

  8. Plutonium Finishing Plant. Interim plutonium stabilization engineering study

    Energy Technology Data Exchange (ETDEWEB)

    Sevigny, G.J.; Gallucci, R.H.; Garrett, S.M.K.; Geeting, J.G.H.; Goheen, R.S.; Molton, P.M.; Templeton, K.J.; Villegas, A.J. [Pacific Northwest Lab., Richland, WA (United States); Nass, R. [Nuclear Fuel Services, Inc. (United States)

    1995-08-01

    This report provides the results of an engineering study that evaluated the available technologies for stabilizing the plutonium stored at the Plutonium Finishing Plant located at the hanford Site in southeastern Washington. Further processing of the plutonium may be required to prepare the plutonium for interim (<50 years) storage. Specifically this document provides the current plutonium inventory and characterization, the initial screening process, and the process descriptions and flowsheets of the technologies that passed the initial screening. The conclusions and recommendations also are provided. The information contained in this report will be used to assist in the preparation of the environmental impact statement and to help decision makers determine which is the preferred technology to process the plutonium for interim storage.

  9. Plutonium Finishing Plant. Interim plutonium stabilization engineering study

    International Nuclear Information System (INIS)

    Sevigny, G.J.; Gallucci, R.H.; Garrett, S.M.K.; Geeting, J.G.H.; Goheen, R.S.; Molton, P.M.; Templeton, K.J.; Villegas, A.J.; Nass, R.

    1995-08-01

    This report provides the results of an engineering study that evaluated the available technologies for stabilizing the plutonium stored at the Plutonium Finishing Plant located at the hanford Site in southeastern Washington. Further processing of the plutonium may be required to prepare the plutonium for interim (<50 years) storage. Specifically this document provides the current plutonium inventory and characterization, the initial screening process, and the process descriptions and flowsheets of the technologies that passed the initial screening. The conclusions and recommendations also are provided. The information contained in this report will be used to assist in the preparation of the environmental impact statement and to help decision makers determine which is the preferred technology to process the plutonium for interim storage

  10. Spray drying of beryllium oxide powder

    International Nuclear Information System (INIS)

    Sepulveda, J.L.; Kahler, D.A.

    1991-01-01

    Forming of beryllia ceramics through dry pressing requires the agglomeration of the powder through spray drying. To produce high quality fired ceramics it is necessary to disperse/grind the primary powder prior to binder addition. Size reduction of the powder is accomplished using an aqueous system in Vibro-Energy mills (VEM) charged with beryllia media to minimize contamination. Two VEM mills of different size were used to characterize the grinding operation. Details of the grinding kinetics are described within the context of the Macroscopic Population Balance Model approach. Spray drying of the ceramic slurry was accomplished with both a centrifugal atomizer and a two fluid nozzle atomizer. Two different spray dryers were used. Important operating parameters affecting the size distribution of the spray dried powder are discussed

  11. A Plutonium Ceramic Target for MASHA

    International Nuclear Information System (INIS)

    Wilk, P A; Shaughnessy, D A; Moody, K J; Kenneally, J M; Wild, J F; Stoyer, M A; Patin, J B; Lougheed, R W; Ebbinghaus, B B; Landingham, R L; Oganessian, Y T; Yeremin, A V; Dmitriev, S N

    2004-01-01

    We are currently developing a plutonium ceramic target for the MASHA mass separator. The MASHA separator will use a thick plutonium ceramic target capable of tolerating temperatures up to 2000 C. Promising candidates for the target include oxides and carbides, although more research into their thermodynamic properties will be required. Reaction products will diffuse out of the target into an ion source, where they will then be transported through the separator to a position-sensitive focal-plane detector array. Experiments on MASHA will allow us to make measurements that will cement our identification of element 114 and provide for future experiments where the chemical properties of the heaviest elements are studied

  12. Studies of plutonium and the lymphatic system: six years of progress at Colorado State University

    International Nuclear Information System (INIS)

    Lebel, J.L.; Bistline, R.W.; Schallberger, J.A.; Dagle, G.E.; Gomez, L.S.

    1976-01-01

    Air-oxidized PuO 2 , high-fired (850 0 C) PuO 2 , Pu(NO 3 ) 4 , PuF 4 , and PuCl 3 , all of which had known amounts of 241 Am, were implanted over the left metacarpus of beagles. The concentration of plutonium and 241 Am in various tissues was measured as a function of time. The effect of diethylenetriaminepentaacetate (DTPA) therapy on organ deposition was studied. Significant differences in the translocation dynamics of air-oxidized PuO 2 , high-fired PuO 2 , and Pu(NO 3 ) 4 in the left superficial cervical lymph node (LSCLN) were observed by in vivo measurement and radiochemical analysis. A 3 percent buildup of plutonium was observed 2 weeks after implantation of air-oxidized PuO 2 , with an exponential buildup to 17 percent at 1 year. With Pu(NO 3 ) 4 and high-fired PuO 2 , the activity accumulation built up to a peak of 31 and 20 percent, respectively, at 30 and 60 days after implantation, followed by a gradual regression. At the implant site 1 year postimplant, a 70 percent reduction in activity was observed with Pu(NO 3 ) 4 , a 90 percent reduction with PuO 2 (high-fired), and only a 20 percent reduction with air-oxidized PuO 2 . The effect of larger particle size on the movement of plutonium from a wound site was speculated. Chemical form and physical form appear to change the rate of translocation and relative depositions of plutonium and americium in various organs.Particle size was not found to be an important factor in the total movement of plutonium in afferent and efferent lymph. More plutonium moves in the cellular fraction of afferent and efferent lymph regardless of particle size. The clearance of plutonium particles from the lymph nodes was associated with necrosis of macrophages. The effect of DTPA therapy on plutonium concentration was greater for the nitrate than for the other chemical forms

  13. Plutonium in a grassland ecosystem

    International Nuclear Information System (INIS)

    Little, C.A.

    1976-01-01

    This study was concerned with plutonium contamination of grassland at the U.S. Energy Research and Development Administration Rocky Flats plant northwest of Denver, Colorado. Of interest were: the definition of major plutonium-containing ecosystem compartments; the relative amounts in those compartments; how those values related to studies done in other geogrphical areas; whether or not the predominant isotopes, 238 Pu and 239 Pu, behaved differently; and what mechanisms might have allowed for the observed patterns of contamination. Samples of soil, litter, vegetation, arthropods, and small mammals were collected for plutonium analysis and mass determination from each of two macroplots. Small aliquots (5 g or less) were analyzed by a rapid liquid scintillation technique and by alpha spectrometry. Of the compartments sampled, greater than 99% of the total plutonium was contained in the soil. The concentrations of plutonium in soil were significantly inversely correlated with distance from the contamination source, depth of the sample, and particle size of the sieved soil samples. The soil data suggested that the distribution of contamination largely resulted from physical transport processes. A mechanism of agglomerated submicron plutonium oxide particles and larger (1-500 μm) host soil particles was proposed. Concentrations of Pu in litter and vegetation were inversely correlated to distance from the source and directly correlated to soil concentrations at the same location. Comparatively high concentration ratios of vegetation to soil suggested wind resuspension of contamination as an important transport mechanism. Arthropod and small mammal samples were highly skewed, kurtotic, and quite variable, having coefficients of variation (standard deviation/mean) as high as 600%. Bone Pu concentrations were lower than other tissues. Hide, GI, and lung were generally not higher in Pu than kidney, liver and muscle

  14. Further development of the ceric oxidation ferrous reduction dichromate titration method of plutonium determination

    International Nuclear Information System (INIS)

    Macdonald, A.; Savage, D.J.

    1986-02-01

    The chemical reactions involved in the amperometric titration method for plutonium have been further examined, to identify remaining potential sources of bias. As these effects were not expected to be more than 0.1% at the 30 mg plutonium level, a procedure for analysing small (3 mg) aliquots of plutonium has been developed, to help identify any marginal effects. At this reduced level both plutonium and 'blank' determinations indicated that reduction of potassium permanganate by oxalic acid is slower than previously suspected and that a variable small bias can result. Decreasing the concentration of the potassium permanganate eliminated the latter effect. A low bias, proportional to the amount of excess oxalic acid added, was observed. Experiments suggest the bias is due to the reaction of oxalate with an intermediary in the iron II/ dichromate titration. The effect can be adequately reduced by complexing oxalate with cerium III. The reduction of plutonium VI by iron II has been shown to be faster than previously assumed, permitting a decreased waiting time before titration. Minor changes to the original procedure are recommended for analysis of 3 mg aliquots of plutonium. (author)

  15. Fabrication of TiC-TiO{sub 2} composite powders by thermal plasma oxidation of titanium carbide powder; Tanka chitan funmatsu no plasma sanka hanno ni yori seiseishita TiC-TiO{sub 2} fukugo funmatsu

    Energy Technology Data Exchange (ETDEWEB)

    Ishigaki, T.; Li, Y.; Haneda, H. [National institute for Research Inorganic materials, Tsukuba (Japan); Kataoka, E. [Showa Cabot Supermetals K.K., Tokyo (Japan)

    2000-09-15

    TiC-TiO{sub 2} composite powders were prepared by in-flight oxidation of titanium carbide powder in RF induction thermal plasmas. Original titanium carbide powder of 20 - 38 {mu}m in particle size was axially injected into the center of argon-oxygen plasma. The powders were partially spheroidized and evaporated through the plasma treatment. X-ray diffraction of plasma-treated powders showed the formation of titanium dioxides, both rutile and anatase phases. The phase content of the plasma-prepared powders strongly depended on the plasma conditions, such as the plasma generating pressure and the oxygen flow rate in plasma generating gas. Especially, the increase of oxygen flow rate in plasma gas gave rise to the increase of heat transfer from plasma to powder particles, exothermic heat of oxidation reaction and cooling rate of plasma, giving the increase of spheroidization ratio, formation ratio of titanium dioxides, and content of anatase phases. (author)

  16. Method of making metal oxide ceramic powders by using a combustible amino acid compound

    Science.gov (United States)

    Pederson, Larry R.; Chick, Lawrence A.; Exarhos, Gregory J.

    1992-01-01

    This invention is directed to the formation of homogeneous, aqueous precursor mixtures of at least one substantially soluble metal salt and a substantially soluble, combustible co-reactant compound, typically an amino acid. This produces, upon evaporation, a substantially homogeneous intermediate material having a total solids level which would support combustion. The homogeneous intermediate material essentially comprises highly dispersed or solvated metal constituents and the co-reactant compound. The intermediate material is quite flammable. A metal oxide powder results on ignition of the intermediate product which combusts same to produce the product powder.

  17. The research of sub-picogram plutonium's quantitative analysis by isotope mass spectrometry

    International Nuclear Information System (INIS)

    Tang Lei; Long Kaiming; Yang Tianli; Liu Xuemei

    2005-10-01

    By using active carbon powder as emission matter, the detect sensitivity of 239 Pu was 0.5 pg (1 ± 0.15, 95% confidence level), improved 20 times compared with the sensitivity 10 pg (1 ± 0.5) of the tradition method. The sub-picogram plutonium sample was determined quantatively by isotope dilution mass spectrometry (IDMS). (authors)

  18. Electroanalytical studies of uranium, neptunium, and plutonium ions in solutions

    International Nuclear Information System (INIS)

    Yoshida, Zenko; Aoyagi, Hisao; Kihara, Sorin

    1989-01-01

    Redox behavior of uranium, neptunium, and plutonium ions, whose oxidation states in acid solutions are between (VI) and (III), were investigated by flow-coulometry with a column electrode of glassy carbon fibers as well as ordinary voltammetry with a rotating disc electrode. Based on current-potential curves the electrode processes were elucidated taking their disproportionation and/or complexation reactions into account. The flow-coulometry, which provides rapid and quantitative electrolysis, was applied to such analytical purposes as follows; the determination of uranium and plutonium in the solution or the solid with discerning their oxidation states, the preparation of species in a desired oxidation state, even in an unstable state which cannot be prepared by ordinary procedure, and the separation of trace amount of uranium in solutions by the electrodeposition of its hydroxide

  19. Supplementation of Citrus maxima Peel Powder Prevented Oxidative Stress, Fibrosis, and Hepatic Damage in Carbon Tetrachloride (CCl4) Treated Rats.

    Science.gov (United States)

    Chowdhury, Mohammed Riaz Hasan; Sagor, Md Abu Taher; Tabassum, Nabila; Potol, Md Abdullah; Hossain, Hemayet; Alam, Md Ashraful

    2015-01-01

    Citrus maxima peel is rich in natural phenolic compounds and has a long use in the traditional medicine. HPLC-DAD analysis on Citrus maxima peel powder exhibited the presence of various phenolic compounds such as caffeic acid and (-)-epicatechin. To determine the plausible hepatoprotective activity of Citrus maxima peel powder, we used carbon tetrachloride (CCl4) treated rat model. Liver damage in rats was confirmed by measuring the AST, ALT, and ALP enzyme activities. In addition, lipid peroxidation products (MDA), nitric oxide, advanced protein oxidation products level (APOP), and catalase activities were also analyzed along with the histological profiling for the inflammatory cell infiltration, collagen, and iron deposition in liver. Dietary supplementation of Citrus maxima peel powder exhibited significant reduction of serum AST, ALT, and ALP activities in carbon tetrachloride treated rats. Moreover, Citrus maxima peel powder also showed a significant reduction of the oxidative stress markers (MDA, NO, and APOP level) and restored the catalase activity in CCl4 treated rats. Histological examination of the liver section revealed reduced inflammatory cells infiltration, collagen, and iron deposition in CCl4 treated rats. The results from this study demonstrated that Citrus maxima peel powder produced significant hepatoprotective action in CCl4 administered rats.

  20. Supplementation of Citrus maxima Peel Powder Prevented Oxidative Stress, Fibrosis, and Hepatic Damage in Carbon Tetrachloride (CCl4 Treated Rats

    Directory of Open Access Journals (Sweden)

    Mohammed Riaz Hasan Chowdhury

    2015-01-01

    Full Text Available Citrus maxima peel is rich in natural phenolic compounds and has a long use in the traditional medicine. HPLC-DAD analysis on Citrus maxima peel powder exhibited the presence of various phenolic compounds such as caffeic acid and (−-epicatechin. To determine the plausible hepatoprotective activity of Citrus maxima peel powder, we used carbon tetrachloride (CCl4 treated rat model. Liver damage in rats was confirmed by measuring the AST, ALT, and ALP enzyme activities. In addition, lipid peroxidation products (MDA, nitric oxide, advanced protein oxidation products level (APOP, and catalase activities were also analyzed along with the histological profiling for the inflammatory cell infiltration, collagen, and iron deposition in liver. Dietary supplementation of Citrus maxima peel powder exhibited significant reduction of serum AST, ALT, and ALP activities in carbon tetrachloride treated rats. Moreover, Citrus maxima peel powder also showed a significant reduction of the oxidative stress markers (MDA, NO, and APOP level and restored the catalase activity in CCl4 treated rats. Histological examination of the liver section revealed reduced inflammatory cells infiltration, collagen, and iron deposition in CCl4 treated rats. The results from this study demonstrated that Citrus maxima peel powder produced significant hepatoprotective action in CCl4 administered rats.

  1. Functional Design Criteria - plutonium stabilization and handling (PUSH) project W-460

    International Nuclear Information System (INIS)

    NELSON, D.W.

    1999-01-01

    This Functional Design Criteria (FDC) contains information to guide the design of the Stabilization and Packaging Equipment necessary to oxidize and package the remaining plutonium-bearing Special Nuclear Materials (SNM) currently in the Plutonium Finishing Plant (PFP) inventory. The FDC also guides the design of vault modifications to allow storage of 3013 packages of stabilized SNM for up to 50 years

  2. Functional Design Criteria plutonium stabilization and handling (PUSH) project W-460

    Energy Technology Data Exchange (ETDEWEB)

    NELSON, D.W.

    1999-09-02

    This Functional Design Criteria (FDC) contains information to guide the design of the Stabilization and Packaging Equipment necessary to oxidize and package the remaining plutonium-bearing Special Nuclear Materials (SNM) currently in the Plutonium Finishing Plant (PFP) inventory. The FDC also guides the design of vault modifications to allow storage of 3013 packages of stabilized SNM for up to 50 years.

  3. Determination of plutonium content in TRR spent fuel by nondestructive neutron counting

    International Nuclear Information System (INIS)

    Chen, Y.-F.; Sheu, R.-J.; Chiao, L.-H.; Yuan, M.-C.; Jiang, S.-H.

    2010-01-01

    For the nuclear safeguard purpose, this work aims to nondestructively determine the plutonium content in the Taiwan Research Reactor (TRR) spent fuel rods in the storage pool before the stabilization process, which transforms the metal spent fuel rods into oxide powder. A SPent-fuel-Neutron-Counter (SPNC) system was designed and constructed to carry out underwater scan measurements of neutrons emitting from the spent fuel rod, from which the 240 Pu mass in the fuel rod will be determined. The SAS2 H control module of the SCALE 5.1 code package was applied to calculate the 240 Pu-to-Pu mass ratio in the TRR spent fuel rod according to the given power history. This paper presents the methodology and design of our detector system as well as the measurements of four TRR spent fuel rods in the storage pool and the comparison of the measured results with the facility declared values.

  4. Study of possible changes brought about by plutonium oxide in the acid phosphatase activity of alveolar macrophages of the rabbit

    International Nuclear Information System (INIS)

    Rouvroy, Huguette

    1970-06-01

    This report describes the various techniques used for determining the acid phosphatase activity of alveolar rabbit macrophages after inhalation of radioactive plutonium oxide particles, exposure of the animals, removal and sampling of the alveolar cells, and technical dosage. The results obtained are presented; they do not make it possible, in this particular case, to affirm that an important change in the enzymatic activity studied occurs. (author) [fr

  5. Influence of cow urine in the bioavailability of plutonium oxide particles in Palomares soils

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa, A.; Aragon, A.; De La Cruz, B.; Gutierrez, J. [CIEMAT, DIAE, Madrid (Spain)

    2004-07-01

    The nuclear accident that occurred in Palomares in 1966 caused the release of plutonium weapon grade particles into a Mediterranean ecosystem, and consequently, urban and farming areas were contaminated with this material. Several studies focussed on the characterization and behaviour of trans-uranides have been carried out in the area. In this work, the solubility evolution of plutonium is analysed for a period of more than 30 years, as well as the influence that the incorporation of cow urine into organic fertilizers has on the solubility of the mentioned element. The average value of the plutonium solubility in water determined in five samples was 0.008% in 1986. However, determinations carried out in samples taken in 1999 and 2000 indicated an increase of the plutonium solubility of 22 to 96 times higher. In order to check the influence of organic fertilizers on the solubility of plutonium, a solubility test was carried out using cow urine as extracting solution. The results show that the solubility of plutonium can reach a value equal to 14%, which is similar to the one obtained with sodium pyrophosphate acting as extracting solution. Thus, these results are a clear warning of what might happen if organic fertilizers are used in transuranic-contaminated soils. (author)

  6. Influence of cow urine in the bioavailability of plutonium oxide particles in Palomares soils

    International Nuclear Information System (INIS)

    Espinosa, A.; Aragon, A.; De La Cruz, B.; Gutierrez, J.

    2004-01-01

    The nuclear accident that occurred in Palomares in 1966 caused the release of plutonium weapon grade particles into a Mediterranean ecosystem, and consequently, urban and farming areas were contaminated with this material. Several studies focussed on the characterization and behaviour of trans-uranides have been carried out in the area. In this work, the solubility evolution of plutonium is analysed for a period of more than 30 years, as well as the influence that the incorporation of cow urine into organic fertilizers has on the solubility of the mentioned element. The average value of the plutonium solubility in water determined in five samples was 0.008% in 1986. However, determinations carried out in samples taken in 1999 and 2000 indicated an increase of the plutonium solubility of 22 to 96 times higher. In order to check the influence of organic fertilizers on the solubility of plutonium, a solubility test was carried out using cow urine as extracting solution. The results show that the solubility of plutonium can reach a value equal to 14%, which is similar to the one obtained with sodium pyrophosphate acting as extracting solution. Thus, these results are a clear warning of what might happen if organic fertilizers are used in transuranic-contaminated soils. (author)

  7. Thermogravimetric study of reduction of oxides present in oxidized nickel-base alloy powders

    Science.gov (United States)

    Herbell, T. P.

    1976-01-01

    Carbon, hydrogen, and hydrogen plus carbon reduction of three oxidized nickel-base alloy powders (a solid solution strengthened alloy both with and without the gamma prime formers aluminum and titanium and the solid solution strengthened alloy NiCrAlY) were evaluated by thermogravimetry. Hydrogen and hydrogen plus carbon were completely effective in reducing an alloy containing chromium, columbium, tantalum, molybdenum, and tungsten. However, with aluminum and titanium present the reduction was limited to a weight loss of about 81 percent. Carbon alone was not effective in reducing any of the alloys, and none of the reducing conditions were effective for use with NiCrAlY.

  8. Characterization of dense lead lanthanum titanate ceramics prepared from powders synthesized by the oxidant peroxo method

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Alexandre H. [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Departamento de Quimica, UFSCar-Universidade Federal de Sao Carlos, Rod.Washington Luis km 235, CP 676 Sao Carlos, SP 13565-905 (Brazil); Souza, Flavio L., E-mail: flavio.souza@ufabc.edu.br [Centro de Ciencias Naturais e Humanas, Universidade Federal do ABC, Rua Santa Adelia 166, Bangu, Santo Andre, SP 09210-170 (Brazil); Chiquito, Adenilson J., E-mail: chiquito@df.ufscar.br [Departamento de Fisica, UFSCar-Federal University of Sao Carlos, Rod.Washington Luis km 235, CP 676 Sao Carlos, SP 13565-905 (Brazil); Longo, Elson, E-mail: elson@iq.unesp.br [Instituto de Quimica de Araraquara, UNESP-Universidade Estadual Paulista, Rua Francisco Degni, CP 355 Araraquara, SP 14801-907 (Brazil); Leite, Edson R., E-mail: derl@power.ufscar.br [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Departamento de Quimica, UFSCar-Universidade Federal de Sao Carlos, Rod.Washington Luis km 235, CP 676 Sao Carlos, SP 13565-905 (Brazil); Camargo, Emerson R., E-mail: camargo@ufscar.br [LIEC-Laboratorio Interdisciplinar de Eletroquimica e Ceramica, Departamento de Quimica, UFSCar-Universidade Federal de Sao Carlos, Rod.Washington Luis km 235, CP 676 Sao Carlos, SP 13565-905 (Brazil)

    2010-12-01

    Nanosized powders of lead lanthanum titanate (Pb{sub 1-x}La{sub x}TiO{sub 3}) were synthesized by means of the oxidant-peroxo method (OPM). Lanthanum was added from 5 to 30% in mol through the dissolution of lanthanum oxide in nitric acid, followed by the addition of lead nitrate to prepare a solution of lead and lanthanum nitrates, which was dripped into an aqueous solution of titanium peroxo complexes, forming a reactive amorphous precipitate that could be crystallized by heat treatment. Crystallized powders were characterized by FT-Raman spectroscopy and X-ray powder diffraction, showing that tetragonal perovskite structure is obtained for samples up to 25% of lanthanum and cubic perovskite for samples with 30% of lanthanum. Powders containing 25 and 30% in mol of lanthanum were calcined at 700 deg. C for 2 h, and in order to determine the relative dielectric permittivity and the phase transition behaviour from ferroelectric-to-paraelectric, ceramic pellets were prepared and sintered at 1100 or 1150 deg. C for 2 h and subjected to electrical characterization. It was possible to observe that sample containing 25% in mol of La presented a normal behaviour for the phase transition, whereas the sample containing 30% in mol of La presented a diffuse phase transition and relaxor behaviour.

  9. Characterization of dense lead lanthanum titanate ceramics prepared from powders synthesized by the oxidant peroxo method

    International Nuclear Information System (INIS)

    Pinto, Alexandre H.; Souza, Flavio L.; Chiquito, Adenilson J.; Longo, Elson; Leite, Edson R.; Camargo, Emerson R.

    2010-01-01

    Nanosized powders of lead lanthanum titanate (Pb 1-x La x TiO 3 ) were synthesized by means of the oxidant-peroxo method (OPM). Lanthanum was added from 5 to 30% in mol through the dissolution of lanthanum oxide in nitric acid, followed by the addition of lead nitrate to prepare a solution of lead and lanthanum nitrates, which was dripped into an aqueous solution of titanium peroxo complexes, forming a reactive amorphous precipitate that could be crystallized by heat treatment. Crystallized powders were characterized by FT-Raman spectroscopy and X-ray powder diffraction, showing that tetragonal perovskite structure is obtained for samples up to 25% of lanthanum and cubic perovskite for samples with 30% of lanthanum. Powders containing 25 and 30% in mol of lanthanum were calcined at 700 deg. C for 2 h, and in order to determine the relative dielectric permittivity and the phase transition behaviour from ferroelectric-to-paraelectric, ceramic pellets were prepared and sintered at 1100 or 1150 deg. C for 2 h and subjected to electrical characterization. It was possible to observe that sample containing 25% in mol of La presented a normal behaviour for the phase transition, whereas the sample containing 30% in mol of La presented a diffuse phase transition and relaxor behaviour.

  10. Dispersion-strengthened Aluminium Products Manufactured by Powder Blending

    DEFF Research Database (Denmark)

    Hansen, Niels

    1969-01-01

    Detailed experiments carried out to examine relationship between microstructure and mechanical properties of powder-blended aluminum products are reported; their results as well as structural studies by transmission electron microscopy and tensile-and creep- testing, are given; as dispersed phase......, various oxide powders were selected on criterion that during manufacturing no reaction must taken place between metal and oxide phase; strength of powder-blended aluminum products increases and elongation decreases with decreasing particle size of aluminum powder and with increasing concentration of oxide...

  11. Evaluation of the Magnesium Hydroxide Treatment Process for Stabilizing PFP Plutonium/Nitric Acid Solutions

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, Mark A.; Schmidt, Andrew J.; Delegard, Calvin H.; Silvers, Kurt L.; Baker, Aaron B.; Gano, Susan R.; Thornton, Brenda M.

    2000-09-28

    This document summarizes an evaluation of the magnesium hydroxide [Mg(OH)2] process to be used at the Hanford Plutonium Finishing Plant (PFP) for stabilizing plutonium/nitric acid solutions to meet the goal of stabilizing the plutonium in an oxide form suitable for storage under DOE-STD-3013-99. During the treatment process, nitric acid solutions bearing plutonium nitrate are neutralized with Mg(OH)2 in an air sparge reactor. The resulting slurry, containing plutonium hydroxide, is filtered and calcined. The process evaluation included a literature review and extensive laboratory- and bench-scale testing. The testing was conducted using cerium as a surrogate for plutonium to identify and quantify the effects of key processing variables on processing time (primarily neutralization and filtration time) and calcined product properties.

  12. High quality aluminium doped zinc oxide target synthesis from nanoparticulate powder and characterisation of sputtered thin films

    Energy Technology Data Exchange (ETDEWEB)

    Isherwood, P.J.M., E-mail: P.J.M.Isherwood@lboro.ac.uk [Centre for Renewable Energy Systems Technology, Loughborough University, Loughborough, Leicestershire LE11 3TU (United Kingdom); Neves, N. [Innovnano, S. A., Rua Coimbra Inovação Parque, IParque Lote 13, 3040-570 Antanhol, Coimbra (Portugal); Bowers, J.W. [Centre for Renewable Energy Systems Technology, Loughborough University, Loughborough, Leicestershire LE11 3TU (United Kingdom); Newbatt, P. [Innovnano, S. A., Rua Coimbra Inovação Parque, IParque Lote 13, 3040-570 Antanhol, Coimbra (Portugal); Walls, J.M. [Centre for Renewable Energy Systems Technology, Loughborough University, Loughborough, Leicestershire LE11 3TU (United Kingdom)

    2014-09-01

    Nanoparticulate aluminium-doped zinc oxide powder was synthesised through detonation and subsequent rapid quenching of metallic precursors. This technique allows for precise compositional control and rapid nanoparticle production. The resulting powder was used to form sputter targets, which were used to deposit thin films by radio frequency sputtering. These films show excellent sheet resistance and transmission values for a wide range of deposition temperatures. Crystal structure analysis shows that crystals in the target have a random orientation, whereas the crystals in the films grow perpendicular to the substrate surface and propagate preferentially along the (002) axis. Higher temperature deposition reduces crystal quality with a corresponding decrease in refractive index and an increase in sheet resistance. Films deposited between room temperature and 300 °C were found to have sheet resistances equivalent to or better than indium tin oxide films for a given average transmission value. - Highlights: • Nanoparticulate AZO powder was used to produce sputter targets. • The powder synthesis technique allows for precise compositional control. • Sputtered films show excellent optical, electronic and structural properties. • High temperature films show reduced electrical and structural quality. • For a given transmission, films show equivalent sheet resistances to ITO.

  13. Conceptual design report, plutonium stabilization and handling,project W-460

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, E.V.

    1997-03-06

    Project W-460, Plutonium Stabilization and Handling, encompasses procurement and installation of a Stabilization and Packaging System (SPS) to oxidize and package for long term storage remaining plutonium-bearing special nuclear materials currently in inventory at the Plutonium Finishing Plant (PFP), and modification of vault equipment to allow storage of resulting packages of stabilized SNM for up to fifty years. This Conceptual Design Report (CDR) provides conceptual design details for the vault modification, site preparation and site interface with the purchased SPS. Two concepts are described for vault configuration; acceleration of this phase of the project did not allow completion of analysis which would clearly identify a preferred approach.

  14. Plutonium and surrogate fission products in a composite ceramic waste form

    International Nuclear Information System (INIS)

    Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T.

    1999-01-01

    Argonne National Laboratory is developing a ceramic waste form to immobilize salt containing fission products and transuranic elements. Preliminary results have been presented for ceramic waste forms containing surrogate fission products such as cesium and the lanthanides. In this work results from scanning electron microscopy/energy dispersive spectroscopy and x-ray diffraction are presented in greater detail for ceramic waste forms containing surrogate fission products. Additionally, results for waste forms containing plutonium and surrogate fission products are presented. Most of the surrogate fission products appear to be silicates or aluminosilicates whereas the plutonium is usually found in an oxide form. There is also evidence for the presence of plutonium within the sodalite phase although the chemical speciation of the plutonium is not known

  15. Evaluation of decontamination during dismantling of plutonium-contaminated glove boxes

    International Nuclear Information System (INIS)

    Kinugasa, Manabu; Taguchi, Seigi; Ohzeki, Satoru; Inoue, Yoshiaki; Kashima, Sadamitsu

    1981-01-01

    The dismantling work of plutonium-contaminated glove boxes was carried out. These glove boxes had been used for the R and D of plutonium-uranium mixed oxide fuel for 15 years. The work was carried out in a pressure-controlled greenhouse, and the contamination of air in the greenhouse was monitored continuously. In order to reduce the contamination of air during dismantling, the decontamination and fixation of loose contaminants on the surfaces of glove boxes were very important. The correlation between decontamination and the contamination of air regarding dismantling is reported in this paper. The surface contamination density of the glove boxes was measured utilizing the smear method before and after the decontamination, and the decontamination effects were estimated. The contamination of air during dismantling was continuously measured with a plutonium dust monitor. It was found that loose contamination exponentially decreased by the decontamination process. When the so-called wet glove boxes, which contained wet recovery and waste disposal apparatus, were dismantled, the contamination of air did not exceed 500 (MPC) a. However, the contamination of air exceeded 500 (MPC) a several times in the present work of dismantling the so-called dry glove boxes which had been used for the fabrication of plutonium-uranium mixed oxide pellets. (Kato, T.)

  16. Electrical conductivity of metal powders under pressure

    Science.gov (United States)

    Montes, J. M.; Cuevas, F. G.; Cintas, J.; Urban, P.

    2011-12-01

    A model for calculating the electrical conductivity of a compressed powder mass consisting of oxide-coated metal particles has been derived. A theoretical tool previously developed by the authors, the so-called `equivalent simple cubic system', was used in the model deduction. This tool is based on relating the actual powder system to an equivalent one consisting of deforming spheres packed in a simple cubic lattice, which is much easier to examine. The proposed model relates the effective electrical conductivity of the powder mass under compression to its level of porosity. Other physically measurable parameters in the model are the conductivities of the metal and oxide constituting the powder particles, their radii, the mean thickness of the oxide layer and the tap porosity of the powder. Two additional parameters controlling the effect of the descaling of the particle oxide layer were empirically introduced. The proposed model was experimentally verified by measurements of the electrical conductivity of aluminium, bronze, iron, nickel and titanium powders under pressure. The consistency between theoretical predictions and experimental results was reasonably good in all cases.

  17. Surplus plutonium disposition draft environmental impact statement. Volume 1, Part A

    International Nuclear Information System (INIS)

    1998-07-01

    On May 22, 1997, DOE published a Notice of Intent (NOI) in the Federal Register (62 Federal Register 28009) announcing its decision to prepare an environmental impact statement (EIS) that would tier from the analysis and decisions reached in connection with the Storage and Disposition of Weapons-Usable Fissile Materials Final Programmatic EIS (Storage and Disposition PEIS). DOE's disposition strategy allows for both the immobilization of surplus plutonium and its use as mixed oxide (MOX) fuel in existing domestic, commercial reactors. The disposition of surplus plutonium would also involve disposal of the immobilized plutonium and MOX fuel (as spent nuclear fuel) in a geologic repository. The Surplus Plutonium Disposition Environmental Impact Statement analyzes alternatives that would use the immobilization approach (for some of the surplus plutonium) and the MOX fuel approach (for some of the surplus plutonium); alternatives that would immobilize all of the surplus plutonium; and the No Action Alternative. The alternatives include three disposition facilities that would be designed so that they could collectively accomplish disposition of up to 50 metric tons (55 tons) of surplus plutonium over their operating lives: (1) the pit disassembly and conversion facility would disassemble pits (a weapons component) and convert the recovered plutonium, as well as plutonium metal from other sources, into plutonium dioxide suitable for disposition; (2) the immobilization facility would include a collocated capability for converting nonpit plutonium materials into plutonium dioxide suitable for immobilization and would be located at either Hanford or SRS. DOE has identified SRS as the preferred site for an immobilization facility; (3) the MOX fuel fabrication facility would fabricate plutonium dioxide into MOX fuel. This volume includes background information; purpose of and need for the proposed action; alternatives for disposition of surplus weapons useable plutonium; and

  18. a Plutonium Ceramic Target for Masha

    Science.gov (United States)

    Wilk, P. A.; Shaughnessy, D. A.; Moody, K. J.; Kenneally, J. M.; Wild, J. F.; Stoyer, M. A.; Patin, J. B.; Lougheed, R. W.; Ebbinghaus, B. B.; Landingham, R. L.; Oganessian, Yu. Ts.; Yeremin, A. V.; Dmitriev, S. N.

    2005-09-01

    We are currently developing a plutonium ceramic target for the MASHA mass separator. The MASHA separator will use a thick plutonium ceramic target capable of tolerating temperatures up to 2000 °C. Promising candidates for the target include oxides and carbides, although more research into their thermodynamic properties will be required. Reaction products will diffuse out of the target into an ion source, where they will then be transported through the separator to a position-sensitive focal-plane detector array. Experiments on MASHA will allow us to make measurements that will cement our identification of element 114 and provide for future experiments where the chemical properties of the heaviest elements are studied.

  19. Effect of Tara (Caesalpinia spinosa Pod Powder on the Oxidation and Colour Stability of Pork Meat Batter During Chilled Storage

    Directory of Open Access Journals (Sweden)

    Monika Skowyra

    2015-01-01

    Full Text Available The eff ect of dried pods of Caesalpinia spinosa, known as tara, on pH, cooking loss, lipid oxidation, colour stability and texture of model meat systems stored at 4 °C for 21 days was investigated. Tara pod powder showing a potential antioxidant activity was added at 0.02, 0.04 and 0.08 % (by mass directly to the pork batt er and compared with a synthetic antioxidant, butylated hydroxyanisole (BHA and control (no added antioxidants. The addition of tara pod powder at 0.02 % was as effective as BHA (0.02 % in retarding lipid oxidation in pork products during storage. Results showed that redness increased after the addition of tara pod powder. Specifically, 0.02 % of tara pod powder was effective in keeping the red colour of meat batter stored under illumination at 4 °C for 48 h. Hardness of pork products was the lowest in samples manufactured with tara pod powder compared with control. Results highlight the potential of using tara pod powder as natural functional ingredient in the development of pork products with enhanced quality and shelf life.

  20. Kinetic study of Cs+ and Eu3+ ions sorption by zirconium oxide powder

    International Nuclear Information System (INIS)

    Hanafi, H.A.; Hassan, H.S.; Hamed, M.M.

    2009-01-01

    Full text: Zirconium oxide powder was chemically synthesized by sol-gel method and characterized using infrared spectra and x-ray diffraction. The sorptive removal of cesium and europium ions from aqueous waste solution using synthetic zirconium oxide powder was investigated using batch technique. Experiments were carried out as a function of pH, time and temperature. The uptake of europium was found to be greater than that of cesium. A comparison of kinetic models applied to the sorption process of each ion was evaluated for the pseudo first order, the pseudo second order, and homogeneous particle diffusion kinetic models, respectively. The results showed that both the pseudo second order and the homogeneous particle diffusion models (HPDM) were found to best correlate the experimental rate data. The numerical values of the rate constants and particle diffusion coefficients were determined from the graphical representation of the proposed models. Activation energy (Ea) and entropy (Δ S*) of activation for each sorption process were also calculated from the linearized form of Arrhenius equation. (author)